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Sample records for advanced thermal hydraulic

  1. Advanced Thermal Hydraulics Design of Commercial SFRs

    International Nuclear Information System (INIS)

    Prototype Fast Breeder Reactor (PFBR) is a 500 MWe pool type sodium cooled fast reactor, which is in an advanced stage of construction in India. As a follow-up to PFBR, six commercial sodium cooled fast reactors (Commercial SFR) of similar capacity are to be constructed, wherein the focus is improved economy and enhanced safety. These reactors are envisaged to have twin-unit concept. Design and construction experiences from PFBR provided the motivation to achieve an optimum design for the Commercial SFR with significant design changes. Some of the changes include, (i) provision of four primary pipes per primary sodium pump, (ii) inner vessel with single torus, (iii) dome shaped roof slab supported on reactor vault, (iv) machined thick plate rotating plugs, (v) reduced main vessel diameter with narrow-gap cooling baffles and (vi) safety vessel integrated with reactor vault. Advanced computational fluid dynamic studies have been performed towards thermal hydraulic design of these components. This paper covers thermal hydraulic design validation of the chosen options, including hot pool thermal hydraulics, influence of control plug shape on pool hydraulics, flow requirement for main vessel cooling, safety analysis of primary pipe rupture event and thermal management top shield and reactor vault. (author)

  2. Thermal hydraulic R and D of Chinese advanced reactors

    International Nuclear Information System (INIS)

    The Chinese government sponsors a program of research, development, and demonstration related to advanced reactors, both small modular reactors and larger systems. These advanced reactors encompass innovative reactor concepts, such as CAP1400 - Chinese large advanced passive pressurized water reactor, Hualong one - Chinese large advanced active and passive pressurized water reactor, ACP100 - Chinese small modular reactor, SCWR- R and D of super critical water-cooled reactor in China, CLEAR - Chinese lead-cooled fast reactor, TMSR - Chinese Thorium molten-salt reactor. The thermal hydraulic R and D of those reactors are summarised. (J.P.N.)

  3. Advanced thermal hydraulic method using 3x3 pin modeling

    International Nuclear Information System (INIS)

    Advanced thermal hydraulic methods are being developed as part of the US DOE sponsored Nuclear Hub program called CASL (Consortium for Advanced Simulation of LWRs). One of the key objectives of the Hub program is to develop a multi-physics tool which evaluates neutronic, thermal hydraulic, structural mechanics and nuclear fuel rod performance in rod bundles to support power uprates, increased burnup/cycle length and life extension for US nuclear plants. Current design analysis tools are separate and applied in series using simplistic models and conservatisms in the analysis. In order to achieve key Nuclear Hub objectives a higher fidelity, multi-physics tool is needed to address the challenge problems that limit current reactor performance. This paper summarizes the preliminary development of a multi-physics tool by performing 3x3 pin modeling and making comparisons to available data. (author)

  4. Equipping simulators with an advanced thermal hydraulics model EDF's experience

    International Nuclear Information System (INIS)

    The development of an accelerated version of the advanced CATHARe-1 thermal hydraulics code designed for EDF training simulators (CATHARE-SIMU) was successfully completed as early as 1991. Its successful integration as the principal model of the SIPA Post-Accident Simulator meant that its use could be extended to full-scale simulators as part of the renovation of the stock of existing simulators. In order to further extend the field of application to accidents occurring in shutdown states requiring action and to catch up with developments in respect of the CATHARE code, EDF initiated the SCAR Project designed to adapt CATHARE-2 to simulator requirements (acceleration, parallelization of the computation and extension of the simulation range). In other respects, the installation of SIPA on workstations means that the authors can envisage the application of this remarkable training facility to the understanding of thermal hydraulics accident phenomena

  5. Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

    International Nuclear Information System (INIS)

    The objectives of the project are to study thermal hydraulic characteristics of advanced nuclear reactor system for evaluating key thermal-hydraulic phenomena relevant to new safety concepts. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. The Followings are main research topics: - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation-induced Thermal Mixing in a Pool - Development of Thermal-Hydraulic Models for Two-Phase Flow - Construction of T-H Data Base

  6. Advanced modelling and numerical strategies in nuclear thermal-hydraulics

    International Nuclear Information System (INIS)

    The first part of the lecture gives a brief review of the current status of nuclear thermal hydraulics as it forms the basis of established system codes like TRAC, RELAP5, CATHARE or ATHLET. Specific emphasis is given to the capabilities and limitations of the underlying physical modelling and numerical solution strategies with regard to the description of complex transient two-phase flow and heat transfer conditions as expected to occur in PWR reactors during off-normal and accident conditions. The second part of the lecture focuses on new challenges and future needs in nuclear thermal-hydraulics which might arise with regard to re-licensing of old plants using bestestimate methodologies or the design and safety analysis of Advanced Light Water Reactors relying largely on passive safety systems. In order to meet these new requirements various advanced modelling and numerical techniques will be discussed including extended wellposed (hyperbolic) two-fluid models, explicit modelling of interfacial area transport or higher order numerical schemes allowing a high resolution of local multi-dimensional flow processes.(author)

  7. Thermal hydraulic evaluation of advanced wire-wrapped assemblies

    International Nuclear Information System (INIS)

    The thermal-hydraulic analyses presented in this report are based on application of the subchannel concept in association with the use of bulk parameters for coolant velocity and coolant temperature within a subchannel. The interactions between subchannels are due to turbulent interchange, pressure-induced diversion crossflow, directed sweeping crossflow induced by the helical wire wrap, and transverse thermal conduction. The FULMIX-II computer program was successfully developed to perform the steady-state temperature predictions for LMFBR fuel assemblies with the reference straight-start design and the advanced wire-wrap designs. Predicted steady-state temperature profiles are presented for a typical CRBRP 217-rod wire-wrapped assembly with the selected wire-wrap designs

  8. NEPTUNE: A new software platform for advanced nuclear thermal hydraulics

    International Nuclear Information System (INIS)

    The NEPTUNE project constitutes the thermal-hydraulic part of the long-term Electricite de France and Commissariat a l'Energie Atomique joint research and development program for the next generation of nuclear reactor simulation tools. This program is also financially supported by the Institut de Radioprotection et Surete Nucleaire and AREVA NP. The project aims at developing a new software platform for advanced two-phase flow thermal hydraulics covering the whole range of modeling scales and allowing easy multi-scale and multidisciplinary calculations. NEPTUNE is a fully integrated project that covers the following fields: software development, research in physical modeling and numerical methods, development of advanced instrumentation techniques, and performance of new experimental programs. The analysis of the industrial needs points out that three main simulation scales are involved. The system scale is dedicated to the overall description of the reactor. The component or subchannel scale allows three-dimensional computations of the main components of the reactors: cores, steam generators, condensers, and heat exchangers. The current generation of system and component codes has reached a very high level of maturity for industrial applications. The third scale, computational fluid dynamics (CFD) in open medium, allows one to go beyond the limits of the component scale for a finer description of the flows. This scale opens promising perspectives for industrial simulations, and the development and validation of the NEPTUNE CFD module have been a priority since the beginning of the project. It is based on advanced physical models (two-fluid or multi field model combined with interfacial area transport and two-phase turbulence) and modern numerical methods (fully unstructured finite volume solvers). For the system and component scales, prototype developments have also started, including new physical models and numerical methods. In addition to scale

  9. Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Song, C. H.; Chung, M. K.; Park, C. K. and others

    2005-04-15

    The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved.

  10. Thermal hydraulic analysis of advanced Pb-Bi cooled NPP using natural circulation

    Science.gov (United States)

    Novitrian, Su'ud, Zaki; Waris, Abdul

    2012-06-01

    We present thermal hydraulic analysis for a low power advanced nuclear reactor cooled by lead-bismuth eutectic. In this work is to study the thermal hydraulic analysis of a low power SPINNOR (Small Power Reactor, Indonesia, No On-site Refuelling) reactor with 125 MWth which a design a core with very small volume and fuel column height, resulting in a negative coolant temperature coefficient and very low channel pressure drop. And also at full power the heat can be completely removed by natural circulation in the primary circuit, thus eliminating the needs for pumps.

  11. Relevant thermal hydraulic aspects of advanced reactors design: status report

    International Nuclear Information System (INIS)

    This status report provides an overview on the relevant thermalhydraulic aspects of advanced reactor designs (e.g. ABWR, AP600, SBWR, EPR, ABB 80+, PIUS, etc.). Since all of the advanced reactor concepts are at the design stage, the information and data available in the open literature are still very limited. Some characteristics of advanced reactor designs are provided together with selected phenomena identification and ranking tables. Specific needs for thermalhydraulic codes together with the list of relevant and important thermalhydraulic phenomena for advanced reactor designs are summarized with the purpose of providing some guidance in development of research plans for considering further code development and assessment needs and for the planning of experimental programs

  12. Strategic Need for Multi-Purpose Thermal Hydraulic Loop for Support of Advanced Reactor Technologies

    Energy Technology Data Exchange (ETDEWEB)

    James E. O' Brien; Piyush Sabharwall; Su-Jong Yoon; Gregory K. Housley

    2014-09-01

    This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation

  13. Thermal-hydraulic Experiments for Advanced Physical Model Development

    International Nuclear Information System (INIS)

    The improvement of prediction models is needed to enhance the safety analysis capability through the fine measurements of local phenomena. To improve the two-phase interfacial area transport model, the various experiments were carried out used SUBO and DOBO. 2x2 and 6x6 rod bundle test facilities were used for the experiment on the droplet behavior. The experiments on the droplet behavior inside a heated rod bundle were focused on the break-up of droplets induced by a spacer grid in a rod bundle geometry. The experiments used GIRLS and JICO and CFD analysis were carried out to comprehend the local condensation of steam jet, turbulent jet induced by condensation and the thermal mixing in a pool. An experimental database of the CHF (Critical Heat Flux) and PDO (Post-dryout) had been constructed. The mechanism of the heat transfer enhancement by surface modifications in nano-fluid was investigated in boiling mode and rapid quenching mode. The special measurement techniques were developed. They are Double -sensor optical void probe, Optic Rod, PIV technique and UBIM system

  14. Thermal-hydraulic Experiments for Advanced Physical Model Development

    International Nuclear Information System (INIS)

    The improvement of prediction models is needed to enhance the safety analysis capability through experimental database of local phenomena. To improve the two-phase interfacial area transport model, the various experiments were carried out with local two-phase interfacial structure test facilities. 2 Χ 2 and 6 Χ 6 rod bundle test facilities were used for the experiment on the droplet behavior. The experiments on the droplet behavior inside a heated rod bundle geometry. The experiments used GIRLS and JICO and CFD analysis were carried out to comprehend the local condensation of steam jet, turbulent jet induced by condensation and the thermal mixing in a pool. In order to develop a model for key phenomena of newly adapted safety system, experiments for boiling inside a pool and condensation in horizontal channel have been performed. An experimental database of the CHF (Critical Heat Flux) and PDO (Post-dryout) was constructed. The mechanism of the heat transfer enhancement by surface modifications in nano-fluid was investigated in boiling mode and rapid quenching mode. The special measurement techniques were developed. They are Double-sensor optical void probe, Optic Rod, PIV technique and UBIM system

  15. Development of an advanced thermal hydraulics model for nuclear power plant simulation

    International Nuclear Information System (INIS)

    This paper summarizes the development of an advanced digital computer thermal hydraulics model for nuclear power plant simulation. A review of thermal hydraulics code design options is presented together with a review of existing engineering models. CAE has developed an unequal temperatures-unequal velocities five equation model based on the drift flux formalism. CAE has selected the model on the basis that phase separation and thermal non-equilibrium are required to simulate complex and important phenomena occurring in systems such as reactor cooling systems (RCS) and steam generators (SG). The drift flux approach to phase separation and countercurrent flow was selected because extensive testing and validation data supports full-range drift flux parameters correlations. The five equation model was also chosen because it conserves important quantities, i.e. mass and energy of each phase, and because of numerical advantages provided by the case of coupling phasic mass conservation equations with phasic energy conservation equations. The basis of CAE's model as well as supporting models for convection and conduction heat transfer, break flow, interphase mass and heat transfer are described. Comparison of code calculations with experimental measurements taken during a small break LOCA test with the OTIS facility are presented. The use of such advanced thermal hydraulics model as plant analyzer considerably improves simulation capabilities of severe transient as well as of normal operation of two phase systems in nuclear power plants. (orig./HP)

  16. Steady-state thermal-hydraulic design analysis of the Advanced Neutron Source reactor

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS) is a research reactor that is planned for construction at Oak Ridge National Laboratory. This reactor will be a user facility with the major objective of providing the highest continuous neutron beam intensities of any reactor in the world. Additional objectives for the facility include providing materials irradiation facilities and isotope production facilities as good as, or better than, those in the High Flux Isotope Reactor. To achieve these objectives, the reactor design uses highly subcooled heavy water as both coolant and moderator. Two separate core halves of 67.6-L total volume operate at an average power density of 4.5 MW(t)/L, and the coolant flows upward through the core at 25 m/s. Operating pressure is 3.1 MPa at the core inlet with a 1.4-MPa pressure drop through the core region. Finally, in order to make the resources available for experimentation, the fuel is designed to provide a 17-d fuel cycle with an additional 4 d planned in each cycle for the refueling process. This report examines the codes and models used to develop the thermal-hydraulic design for ANS, as well as the correlations and physical data; evaluates thermal-hydraulic uncertainties; reports on thermal-hydraulic design and safety analysis; describes experimentation in support of the ANS reactor design and safety analysis; and provides an overview of the experimental plan

  17. Selected thermal and hydraulic experimentation in support of the advanced neutron source reactor

    International Nuclear Information System (INIS)

    The ANS Reactor has unique thermal-hydraulic characteristics in comparison to other research and commercial reactors: Heavy water coolant, Parallel Rectangular channels (involute), Very small channel gap (1.27 mm), Very high velocity (25 m/s), Very high exit subcooling, Moderately high heat flux, High average power density. The objective was to determine experimentally the appropriate core thermal hydraulic limits at ANS conditions. Advanced Neutron Source (ANS) Thermal Hydraulic Test Loop (THTL) was designed to operate in 'Stiff', 'Soft' and 'Modified Stiff' Modes.Summary of Thermal Hydraulic Limit Testing and Analysis shows: FE data has been acquired at ANS typical flow velocities; An extensive OSV/OFI data base has been developed with a very broad parameter range, A modification of the Saha-Zuber correlation was proposed to account for reduced subcooling effects; Closeout activities include continued investigation of wider span test channels; Some testing for HFIR will be performed to evaluate the effect of reduced channel gap; Future plans called for additional testing at 3-core conditions, hot spot testing, etc. The Objective of Fuel Plate Stability Testing was to experimentally evaluate the structural response of ANS fuel plates to hydraulic loads. Summary of Fuel Plate Stability Testing shows: A Method Has Been Developed to Predict Structural Response of Fuel Plates to Hydraulic Loading Prediction of AP across plates Determine deflection/stress levels using structural analysis; ANS, Specific Conclusions are: no evidence of potential plate collapse in the coolant velocity range from 050 m/s, no evidence of plate flutter with coolant velocities below 33 m/s, local stress levels appear to dictate plate limits as opposed to plate deflection. The objective of Flow Blockage Testing was to experimentally determine local thermal and fluid. Summary of Flow Blockage Testing and Analysis showed: CFD code has been benchmarked against prototypic ANS flow conditions and

  18. Advanced thermal-hydraulic and neutronic codes: current and future applications. Summary and conclusions

    International Nuclear Information System (INIS)

    An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for

  19. Proceedings of the workshop on advanced thermal-hydraulic and neutronic codes: current and future applications

    International Nuclear Information System (INIS)

    An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for

  20. Update to advanced neutron source steady-state thermal-hydraulic report

    International Nuclear Information System (INIS)

    This report is intended to be a supplement to ORNL/TM-12398, Steady-State Thermal-Hydraulic Design Analysis of the Advanced Neutron Source Reactor. It updates the core thermal-hydrualic design to the latest three-element configuration and also provides the most recent information on the thermal-hydraulic statistical uncertainty analysis. In addition, it includes calculations of beam tube cooling and control rod lift forces, which were not addressed in the initial report. This report describes work that is a snapshot in time as it stood at the end of the project. The three-element core calculations include a description of changes made to the overall coolant system; however, most of the analysis is focused on fuel loading thermal-hydraulic calculations. This analysis uses updated uncertainty values and indicates that a two-dimensional fuel grading in the three-element core would still be necessary to meet the desired operating and safety criteria. Analysis of cooling in the reflector tank examines various cooling options for the reflector tank components. This work investigated multiple forced convection designs as well as natural convection cooling requirements. Lift forces on the inner control rods caused by the upward coolant flow were also examined. Initial control rod designs were such that a sheared control rod would tend to lift because of flow forces. Design changes were recommended that would eliminate this issue. They included geometry changes to the inner control rod cooling channels, changes to the orificing in the central hole region, and reduction of inner control rod coolant velocity

  1. Thermal-hydraulic studies of the Advanced Neutron Source cold source

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS), in its conceptual design phase at Oak Ridge National Laboratory, was to be a user-oriented neutron research facility producing the most intense steady-state flux of thermal and cold neutrons in the world. Among its many scientific applications, the production of cold neutrons was a significant research mission for the ANS. The cold neutrons come from two independent cold sources positioned near the reactor core. Contained by an aluminum alloy vessel, each cold source is a 410-mm-diam sphere of liquid deuterium that functions both as a neutron moderator and a cryogenic coolant. With nuclear heating of the containment vessel and internal baffling, steady-state operation requires close control of the liquid deuterium flow near the vessel's inner surface. Preliminary thermal-hydraulic analyses supporting the cold source design were performed with heat conduction simulations of the vessel walls and multidimensional computational fluid dynamics simulations of the liquid deuterium flow and heat transfer. This report presents the starting phase of a challenging program and describes the cold source conceptual design, the thermal-hydraulic feasibility studies of the containment vessel, and the future computational and experimental studies that were planned to verify the final design

  2. Thermal-hydraulics numerical analyses of Pebble Bed Advanced High Temperature Reactor hot channel

    International Nuclear Information System (INIS)

    Background: The thermal hydraulics behavior of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) hot channel was studied. Purpose: We aim to analyze the thermal-hydraulics behavior of the PB-AHTR, such as pressure drop, temperature distribution of coolant and pebble bed as well as thermal removal capacity in the condition of loss of partial coolant. Methods: We used a modified FLUENT code which was coupled with a local non-equilibrium porous media model by introducing a User Defined Scalar (UDS) in the calculation domain of the reactor core and subjoining different resistance terms (Ergun and KTA) to calculate the temperature of coolant, solid phase of pebble bed and pebble center in the core. Results: Computational results showed that the resistance factor has great influence on pressure drop and velocity distribution, but less impact on the temperature of coolant, solid phase of pebble bed and pebble center. We also confirmed the heat removal capacity of the PB-AHTR in the condition of nominal and loss of partial coolant conditions. Conclusion: The numerical analyses results can provide a useful proposal to optimize the design of PB-AHTR. (authors)

  3. Development of a steady thermal-hydraulic analysis code for the China Advanced Research Reactor

    Institute of Scientific and Technical Information of China (English)

    TIAN Wenxi; QIU Suizheng; GUO Yun; SU Guanghui; JIA Dounan; LIU Tiancai; ZHANG Jianwei

    2007-01-01

    A multi-channel model steady-state thermalhydraulic analysis code was developed for the China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed mass flow distribution in the core was obtained. The result shows that structure size plays the most important role in mass flow distribution, and the influence of core power could be neglected under singlephase flow. The temperature field of the fuel element under unsymmetrical cooling condition was also obtained, which is necessary for further study such as stress analysis, etc. Of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of the mean and hot channel was carried out and it is proved that all thermal-hydraulic parameters satisfy the "Safety design regulation of CARR".

  4. Thermally Actuated Hydraulic Pumps

    Science.gov (United States)

    Jones, Jack; Ross, Ronald; Chao, Yi

    2008-01-01

    Thermally actuated hydraulic pumps have been proposed for diverse applications in which direct electrical or mechanical actuation is undesirable and the relative slowness of thermal actuation can be tolerated. The proposed pumps would not contain any sliding (wearing) parts in their compressors and, hence, could have long operational lifetimes. The basic principle of a pump according to the proposal is to utilize the thermal expansion and contraction of a wax or other phase-change material in contact with a hydraulic fluid in a rigid chamber. Heating the chamber and its contents from below to above the melting temperature of the phase-change material would cause the material to expand significantly, thus causing a substantial increase in hydraulic pressure and/or a substantial displacement of hydraulic fluid out of the chamber. Similarly, cooling the chamber and its contents from above to below the melting temperature of the phase-change material would cause the material to contract significantly, thus causing a substantial decrease in hydraulic pressure and/or a substantial displacement of hydraulic fluid into the chamber. The displacement of the hydraulic fluid could be used to drive a piston. The figure illustrates a simple example of a hydraulic jack driven by a thermally actuated hydraulic pump. The pump chamber would be a cylinder containing encapsulated wax pellets and containing radial fins to facilitate transfer of heat to and from the wax. The plastic encapsulation would serve as an oil/wax barrier and the remaining interior space could be filled with hydraulic oil. A filter would retain the encapsulated wax particles in the pump chamber while allowing the hydraulic oil to flow into and out of the chamber. In one important class of potential applications, thermally actuated hydraulic pumps, exploiting vertical ocean temperature gradients for heating and cooling as needed, would be used to vary hydraulic pressures to control buoyancy in undersea research

  5. Preliminary research on RTDP methodology for advanced LPP thermal-hydraulic design

    International Nuclear Information System (INIS)

    Departure from nucleate boiling (DNB) design basis is one of the most important basis for reactor core thermal-hydraulic design. In order to evaluate whether the DNB design basis meets the demand of thermal-hydraulic design, the departure from nucleate boiling ratio (DNBR) design limit should be determined first. The RTDP methodology was described detailedly, in which the uncertainties of operating parameters and nuclear design parameters were statistically combined. Then the RTDP methodology and a reactor subchannel code were applied to calculate the DNBR design limit and quality limit for LPP. The conclusions were presented to provide the key acceptable criterion for DNBR design basis. (authors)

  6. Thermal hydraulic test apparatus to develop advanced BWR fuel bundles with spectral shift rods (SSR)

    International Nuclear Information System (INIS)

    An advanced water rod (WR) called the spectral shift rod (SSR), which replaces a conventional WR in a BWR fuel bundle, enhances the BWR's merit of uranium saving through the spectral shift operation. The SSR consists of an inlet hole, a wide ascending path, a narrow descending path and an outlet hole. The inlet hole locates below a lower tie plate (LTP) and the outlet hole is set above it. In the SSR, water boils by neutron and gamma-ray heating and water level is formed in the ascending path. This SSR water level can be controlled by core flow rate, which amplifies core void fraction change, resulting in the amplified spectral shift effect. Steady state and transient tests were conducted to evaluate SSR thermal-hydraulic characteristics under BWR operation condition. The several types of SSR configuration were tested, which covers SSR design in both next generation and conventional BWRs. In this paper, the test apparatus overview and measurement systems especially two phase water level measures in the SSR are presented. (author)

  7. Korean development of advanced thermal-hydraulic codes for water reactors and HTGRS: space and gamma

    International Nuclear Information System (INIS)

    Korea has been developing SPACE(Safety and Performance Analysis CodE) and GAMMA(GAs Multicomponent Mixture Analysis) codes for safety analysis of PWRs and HTGRs, respectively. SPACE is being developed by the Korea nuclear industry, which is a thermal-hydraulic analysis code for safety analysis of a PWR. It will replace outdated vendor supplied codes and will be used for the safety analysis of operating PWR and the design of an advanced PWR. It consists of the up-to-date physical models of two-phase flow dealing with multi-dimensional two-fluid, three-field flow. The GAMMA code consists of the multi-dimensional governing equations consisting of the basic equations for continuity, momentum conservation, energy conservation of the gas mixture, and mass conservation of n species. GAMMA is based on a porous media model so that we can deal with the thermo-fluid and chemical reaction behaviors in a multicomponent mixture system as well as heat transfer within the solid components, free and forced convection between a solid and a fluid, and radiative heat transfer between the solid surfaces. GAMMA has a model for helium turbines for HTGRs based on the throughflow calculation. We performed extensive code assessment for the V&V of SPACE and GAMMA. (author)

  8. Thermal-hydraulics analysis for advanced fuel to be used in Candu 600 nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Catana, Alexandru [RAAN, Institute for Nuclear Research, Str. Campului Nr. 1, Pitesti, Arges (Romania); Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel [University POLITEHNICA of Bucharest (Romania)

    2008-07-01

    Two Candu 600 pressure tube nuclear reactors cover about 17% of Romania's electricity demand. These nuclear reactors are moderated/cooled with D{sub 2}O, fuelled on-power with Natural Uranium (NU) dioxide encapsulated in a standard (STD37) fuel bundle. High neutron economy is achieved using D{sub 2}O as moderator and coolant in separated systems. To reduce fuel cycle costs, programs were initiated in Canada, S.Korea, Argentina and Romania for the design and build new fuel bundles able to accommodate different fuel compositions. Candu core structure and modular fuel bundles, permits flexible fuel cycles. The main expected achievements are: reduced fuel cycle costs, increased discharge burn-up, plutonium and minor actinides management, thorium cycle, use of recycled PWR and in the same time waste minimization and operating cost reduction. These new fuel bundles are to be used in already operated Candu reactors. Advanced fuel bundle were proposed: CANFLEX bundle (Canada, S-Korea); the Romanian 'SEU43' bundle (Fig 1). In this paper thermal-hydraulic analysis in sub-channel approach is presented for SEU43. Comparisons with standard (STD37) fuel bundles are made using SEU-NU for NU fuel composition and SEU-0.96, for recycled uranium (RU) fuel with 0.96% U-235. Extended and comprehensive analysis must be made in order to assess the TH behaviour of SEU43. In this paper, considering STD37, SEU43-NU and SEU43-0.96 fuel bundles, main TH parameters were analysed: pressure drop, fuel highest temperatures, coolant density, critical heat flux. Differences between these fuel types are outlined. Benefits are: fuel costs reduction, spent fuel waste minimization, increase in competitiveness of nuclear power. Safety margins must be, at least, conserved. (authors)

  9. Validation of thermal hydraulic computer codes for advanced light water reactor

    International Nuclear Information System (INIS)

    The Czech Republic operates 4 WWER-440 units, two WWER-1000 units are being finalised (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppressure system are modelled with RALOC and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems. An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. The paper provides a concise information on these activities of the NRI and its Thermal-hydraulics Department. A detailed example of the system code validation and the consequent utilisation of the results for a real NPP purposes is included. (author)

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

    International Nuclear Information System (INIS)

    The main objective of the Thermal Hydraulic (TH) analysis of reactor core and fuel assembly design is the determination of pressure loss and critical heat flux (CHF). Especially the description of the latter effect requires the modeling of a large variety of physical phenomena starting with single phase quantities like turbulence or fluid-wall friction, two phase quantities like void distributions, heat transfer between fuel rod and fluid and ultimately the CHF mechanism itself. Additional complexity is added by the fact that the relevant geometric scales which have to be resolved, cover a wide range from the length of the fuel assembly (∼ 4000 mm), over the typical dimensions of sub-channel cross sections and the vanes on the spacer grids (∼ 10 mm) down to the microscopic scales set by bubble sizes and boundary layers (mm to sub mm). Due to the above described situation the necessary TH quantities are often determined by measurements. The main advantage of this technique is that measurements are widely accepted and trusted if the geometry and flow conditions are sufficiently close to real reactor conditions. The main disadvantage of experiments is that they are expensive both with respect to time and money; especially in high pressure tests they give only limited access to the test object. Consequently there is a strong interest to develop computer codes with the goal of minimizing the need of experiments, and hence, speeding up and reducing costs of fuel assembly and core design. Today most of the design work is based on sub-channel codes, originally developed in the 70's; they provide an effective description of the TH in fuel assemblies by regarding the fuel assembly as a system of communicating channels (the volume enclosed by four fuel rods = one sub-channel). Further development of these codes is one main focus of AREVA NP's Thermal Hydraulic method and code development strategy. To focus the know-how and resources existing in the different regions of

  11. Power generation costs and ultimate thermal hydraulic power limits in hypothetical advanced designs with natural circulation

    International Nuclear Information System (INIS)

    Maximum power limits for hypothetical designs of natural circulation plants can be described analytically. The thermal hydraulic design parameters are those which limit the flow, being the elevations, flow areas, and loss coefficients. WE have found some simple ''design'' equations for natural circulation flow to power ratio, and for the stability limit. The analysis of historical and available data for maximum capacity factor estimation shows 80% to be reasonable and achievable. The least cost is obtained by optimizing both hypothetical plant performance for a given output,a nd the plant layout and design. There is also scope to increase output and reduce cost by considering design variations of primary and secondary pressure, and by optimizing component elevations and loss coefficients. The design limits for each are set by stability and maximum flow considerations, which deserve close and careful evaluation

  12. Thermal-hydraulic experiments of an advanced PIUS-type reactor

    International Nuclear Information System (INIS)

    The author constructed a semi-large scale experimental apparatus for simulating thermal-hydraulic behavior of the PIUS-type reactor with keeping the volumetric scaling ratio to the realistic reactor model. Fundamental experiments such as a steady state operation and a pump trip simulation were reported in ICONE-3(1995). In this paper the authors present two main results. One is a feedback control system using the upper density lock, and a start up simulation based on the non-uniform heating for both the primary loop and the poison loop. The other is a control system of small scale sub-loop attached to the poison loop in order to establish PIUS principle on the realistic operation of the PIUS-type reactor

  13. Advances of study on thermal-hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    R and D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) is started at Japan Atomic Energy Research Institute in collaboration with power company, reactor vendors, universities since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R and D items for the RMWR because of the tight-lattice configuration. In this paper, we will show the R and D plan and describe some advances on experimental and analytical studies. The experimental study is performed mainly using large-scale (37-rod bundle) test facility and the analytical one aims to develop a predictable technology for geometry effects such as gap between rods, grid spacer configuration etc. using advanced 3-D two-phase flow simulation methods. Steady-state and transient critical power experiments are conducted with the test facility (Gap width between rods: 1.0 mm) and the experimental data reveal the feasibility of RMWR. (authors)

  14. A review of modern advances in analyses and applications of single-phase natural circulation loop in nuclear thermal hydraulics

    International Nuclear Information System (INIS)

    Highlights: • Comprehensive review of state-of-the-art on single-phase natural circulation loops. • Detailed discussion on growth in solar thermal system and nuclear thermal hydraulics. • Systematic development in scaling methodologies for fabrication of test facilities. • Importance of numerical modeling schemes for stability assessment using 1-D codes. • Appraisal of current trend of research and possible future directions. - Abstract: A comprehensive review of single-phase natural circulation loop (NCL) is presented here. Relevant literature reported since the later part of 1980s has been meticulously surveyed, with occasional obligatory reference to a few pioneering studies originating prior to that period, summarizing the key observations and the present trend of research. Development in the concept of buoyancy-induced flow is discussed, with introduction to flow initiation in an NCL due to instability. Detailed discussion on modern advancement in important application areas like solar thermal systems and nuclear thermal hydraulics are presented, with separate analysis for various reactor designs working on natural circulation. Identification of scaling criteria for designing lab-scale experimental facilities has gone through a series of modification. A systematic analysis of the same is presented, considering the state-of-the-art knowledge base. Different approaches have been followed for modeling single-phase NCLs, including simplified Lorenz system mostly for toroidal loops, 1-D computational modeling for both steady-state and stability characterization and 3-D commercial system codes to have a better flow visualization. Methodical review of the relevant studies is presented following a systematic approach, to assess the gradual progression in understanding of the practical system. Brief appraisal of current research interest is reported, including the use of nanofluids for fluid property augmentation, marine reactors subjected to rolling waves

  15. 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

  16. Thermal-hydraulic unreliability of passive systems

    International Nuclear Information System (INIS)

    Advanced light water reactor designs like AP600 and the simplified boiling water reactor (SBWR) use passive safety systems for accident prevention and mitigation. Because these systems rely on natural forces for their operation, their unavailability due to hardware failures and human error is significantly smaller than that of active systems. However, the coolant flows predicted to be delivered by these systems can be subject to significant uncertainties, which in turn can lead to a significant uncertainty in the predicted thermal-hydraulic performance of the plant under accident conditions. Because of these uncertainties, there is a probability that an accident sequence for which a best estimate thermal-hydraulic analysis predicts no core damage (success sequence) may actually lead to core damage. For brevity, this probability will be called thermal-hydraulic unreliability. The assessment of this unreliability for all the success sequences requires very expensive computations. Moreover, the computational cost increases drastically as the required thermal-hydraulic reliability increases. The required computational effort can be greatly reduced if a bounding approach can be used that either eliminates the need to compute thermal-hydraulic unreliabilities, or it leads to the analysis of a few bounding sequences for which the required thermal-hydraulic reliability is relatively small. The objective of this paper is to present such an approach and determine the order of magnitude of the thermal-hydraulic unreliabilities that may have to be computed

  17. Advanced Multiphysics Thermal-Hydraulics Models for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Prashant K [ORNL; Freels, James D [ORNL

    2015-01-01

    Engineering design studies to determine the feasibility of converting the High Flux Isotope Reactor (HFIR) from using highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL). This work is part of an effort sponsored by the US Department of Energy (DOE) Reactor Conversion Program. HFIR is a very high flux pressurized light-water-cooled and moderated flux-trap type research reactor. HFIR s current missions are to support neutron scattering experiments, isotope production, and materials irradiation, including neutron activation analysis. Advanced three-dimensional multiphysics models of HFIR fuel were developed in COMSOL software for safety basis (worst case) operating conditions. Several types of physics including multilayer heat conduction, conjugate heat transfer, turbulent flows (RANS model) and structural mechanics were combined and solved for HFIR s inner and outer fuel elements. Alternate design features of the new LEU fuel were evaluated using these multiphysics models. This work led to a new, preliminary reference LEU design that combines a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone. Preliminary results of estimated thermal safety margins are presented. Fuel design studies and model enhancement continue.

  18. Nonelastomeric Rod Seals for Advanced Hydraulic Systems

    Science.gov (United States)

    Hady, W. F.; Waterman, A. W.

    1976-01-01

    Advanced high temperature hydraulic system rod sealing requirements can be met by using seals made of nonelastomeric (plastic) materials in applications where elastomers do not have adequate life. Exploratory seal designs were optimized for advanced applications using machinable polyimide materials. These seals demonstrated equivalent flight hour lives of 12,500 at 350 F and 9,875 at 400 F in advanced hydraulic system simulation. Successful operation was also attained under simulated space shuttle applications; 96 reentry thermal cycles and 1,438 hours of vacuum storage. Tests of less expensive molded plastic seals indicated a need for improved materials to provide equivalent performance to the machined seals.

  19. International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Highlights: • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. - Abstract: In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety

  20. International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hainoun, A., E-mail: pscientific2@aec.org.sy [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400 S.C de Bariloche, Rio Negro (Argentina); Umbehaun, P. [Centro de Engenharia Nuclear – CEN, IPEN-CNEN/SP, Av. Lineu Prestes 2242-Cidade Universitaria, CEP-05508-000 São Paulo, SP (Brazil); Chatzidakis, S. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Ghazi, N. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Park, S. [Research Reactor Design and Engineering Division, Basic Science Project Operation Dept., Korea Atomic Energy Research Institute (Korea, Republic of); Mladin, M. [Institute for Nuclear Research, Campului Street No. 1, P.O. Box 78, 115400 Mioveni, Arges (Romania); Shokr, A. [Division of Nuclear Installation Safety, Research Reactor Safety Section, International Atomic Energy Agency, A-1400 Vienna (Austria)

    2014-12-15

    Highlights: • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. - Abstract: In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety

  1. 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

  2. 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

  3. Advanced Performance Hydraulic Wind Energy

    Science.gov (United States)

    Jones, Jack A.; Bruce, Allan; Lam, Adrienne S.

    2013-01-01

    The Jet Propulsion Laboratory, California Institute of Technology, has developed a novel advanced hydraulic wind energy design, which has up to 23% performance improvement over conventional wind turbine and conventional hydraulic wind energy systems with 5 m/sec winds. It also has significant cost advantages with levelized costs equal to coal (after carbon tax rebate). The design is equally applicable to tidal energy systems and has passed preliminary laboratory proof-of-performance tests, as funded by the Department of Energy.

  4. Thermal hydraulic studies for passive heat transport systems relevant to advanced reactors

    International Nuclear Information System (INIS)

    Nuclear is the only non-green house gas generating power source that can replace fossil fuels and can be commercially deployed in large scale. However, the enormous developmental efforts and safety upgrades during the past six decades have somewhat eroded the economic competitiveness of water-cooled reactors which form the mainstay of the current nuclear power programme. Further, the introduction of the supercritical Rankine cycle and the gas turbine based advanced fuel cycles have enhanced the efficiency of fossil fired power plants (FPP) thereby reducing its greenhouse gas emissions. The ongoing development of ultra-supercritical and advanced ultra-supercritical turbines aims to further reduce the greenhouse gas emissions and economic competitiveness of FPPs. In the backdrop of these developments, the nuclear industry also initiated development of advanced nuclear power plants (NPP) with improved efficiency, sustainability and enhanced safety as the main goals. A review of the advanced reactor concepts being investigated currently reveals that excepting the SCWR, all other concepts use coolants other than water. The coolants used are lead, lead bismuth eutectic, liquid sodium, molten salts, helium and supercritical water. Besides, some of these are employing passive systems to transport heat from the core under normal operating conditions. In view of this, a study is in progress at BARC to examine the performance of simple passive systems using SC CO2, SCW, LBE and molten salts as the coolant. This paper deals with some of the recent results of these studies. The study focuses on the steady state, transient and stability behaviour of the passive systems with these coolants. (author)

  5. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    International Nuclear Information System (INIS)

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors

  6. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [Brookhaven National Lab., Upton, NY (United States)] [comp.

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.

  7. Consideration of a design optimization method for advanced nuclear power plant thermal-hydraulic components

    International Nuclear Information System (INIS)

    In order to meet the global energy demand and also mitigate climate change, we anticipate a significant resurgence of nuclear power in the next 50 years. Globally, Generation III plants (ABWR) have been built; Gen' III+ plants (EPR, AP1000 others) are anticipated in the near term. The U.S. DOE and Japan are respectively pursuing the NGNP and MSFR. There is renewed interest in closing the fuel cycle and gradually introducing the fast reactor into the LWR-dominated global fleet. In order to meet Generation IV criteria, i.e. thermal efficiency, inherent safety, proliferation resistance and economic competitiveness, plant and energy conversion system engineering design have to increasingly meet strict design criteria with reduced margin for reliable safety and uncertainties. Here, we considered a design optimization approach using an anticipated NGNP thermal system component as a Case Study. A systematic, efficient methodology is needed to reduce time consuming trial-and-error and computationally-intensive analyses. We thus developed a design optimization method linking three elements; that is, benchmarked CFD used as a 'design tool', artificial neural networks (ANN) to accommodate non-linear system behavior and enhancement of the 'design space', and finally, response surface methodology (RSM) to optimize the design solution with targeted constraints. The paper presents the methodology including guiding principles, an integration of CFD into design theory and practice, consideration of system non-linearities (such as fluctuating operating conditions) and systematic enhancement of the design space via application of ANN, and a stochastic optimization approach (RSM) with targeted constraints. Results from a Case Study optimizing the printed circuit heat exchanger for the NGNP energy conversion system will be presented. (author)

  8. Thermal/hydraulic computer program

    International Nuclear Information System (INIS)

    This technical paper describes a user-friendly computer program designed to analyze fluid-piping systems. Both the thermal and the hydraulic aspects of the problem are addressed simultaneously in order to provide coupled solutions. The program was developed specifically for aerospace applications such as the waste-heat acquisition and transport loops on the Space Station; however, the code is general and could be applied readily to any fluid loop. The thermal portion of the program is an existing analyzer, Mini-MITAS. This PC version of the generalized thermal analyzer MITAS can handle up to 1000 nodes and uses the traditional RC network approach for solving the thermal portion of the problem. Since this type of thermal analyzer is well known in the thermal community (BETA, SINDA, CINDA, and MITAS), the emphasis in this paper is on the hydraulic portion of the program

  9. Thermal hydraulic tests for reactor safety system -Research on the improvement of nuclear safety-

    International Nuclear Information System (INIS)

    The present research aims at the development of the thermal hydraulic verification test technology for the reactor safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. (Author)

  10. Proceedings of the ANS/ASME/NRC international topical meeting on nuclear reactor thermal-hydraulics: LMFBR and HTGR advanced reactor concepts and analysis methods

    International Nuclear Information System (INIS)

    Separate abstracts are included for each of the papers presented concerning the thermal-hydraulics of LMFBR type reactors; mathematical methods in nuclear reactor thermal-hydraulics; heat transfer in gas-cooled reactors; and thermal-hydraulics of pebble-bed reactors. Two papers have been previously abstracted and input to the data base

  11. Thermal hydraulic analysis due to the changes in heat removal for advanced heavy water reactor

    International Nuclear Information System (INIS)

    Advanced Heavy Water Reactor is a natural circulation light water cooled and heavy water moderated pressure tube reactor. Changes in heat removal by primary heat transport system of the reactor have significant impact on various important system parameters like pressures, qualities, reactor power and flows. Increase in heat removal leads to the cooldown of the system subsequently reducing pressure, void increase and changes in power and flows of the system. Decrease in heat removal leads to warm-up of the system subsequently raising pressure, void collapse, and changes in power and flows of the system. The behaviour is complex as system under consideration is natural circulation system. This article presents the results of simulations made with the RELAP5-MOD3.2 code that show first the impact of a decrease in feed water temperature on fluid temperature, steam drum pressure, core exit void, reactivity, reactor power, core flow, steam flow and clad temperature and secondly the impact of a loss of normal feed water flow on steam drum pressure, channel flow, core quality, clad surface temperature. For lowering of feed water temperature transient and in isolation condensers cold water injection, the reactor power increases and the reactor trips on the high power signal. Simultaneous flow increment due to the 2 phase natural circulation characteristic has caused the clad temperature to limit to their steady state value. In case of loss of feed transient the reactor trips on high pressure. The clad surface temperature rise from steady state operating value is marginal and it is well within the safety limit as per the acceptance criteria

  12. KATHY: Framatome ANP's thermal hydraulic test loop

    International Nuclear Information System (INIS)

    The investigation of the thermal-hydraulic behavior of fuel assemblies (FA) under simulated reactor conditions will continue to be a key component of fuel assembly design and development work in the future. Today's fuel assemblies are highly complex. Optimum in-pile performance is only assured, if each and every component part is carefully matched to the others in the reactor core. Since, even today, it is a challenge to mathematically simulate with the necessary degree of accuracy the thermal-hydraulic processes occurring while reactor coolant is flowing through the fuel assemblies, experimental validation of empirical analyses is especially important. Additionally, data gained from the performance of thermal hydraulic tests under realistic reactor conditions are indispensable for the qualification and validation of codes and methods, which are used to estimate the extended operation limits of advanced boiling water reactor (BWR) and pressurized water reactor (PWR) FA. That's why the operation of a validated, state-of-the-art thermal-hydraulic test facility is an absolute must for Framatome ANP. Framatome ANP's Multi-Function Thermal-Hydraulic Test Loop KATHY in Germany (KArlstein Thermal HYdraulics) has been in operation since 1986. KATHY is continually being improved and expanded at great expense. KATHY is not only used for measuring the critical heat flux (CHF) of BWR and PWR FA, but also for 1 and 2-phase flow measurements, for void-, plant transient- and stability measurements. So far, more than 15,000 tests have successfully been conducted on numerous BWR and PWR bundle designs. (author)

  13. Thermal-Hydraulics Research in the Valencia Polytechnic University

    International Nuclear Information System (INIS)

    The research on thermal-hydraulics at the Polytechnic University of Valencia is performed by the TIN group (thermal-hydraulic and Nuclear Engineering). The group activities are currently carried out at the Energy Engineering Institute. The main research topics are: transient analysis of reactors, nuclear reactor stability, passive and advanced safety reactors, two-phase flow in nuclear reactors. (Author)

  14. Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles: III - Numerical Evaluation of Fluid Mixing Phenomena using Advanced Interface-Tracking Method -

    Science.gov (United States)

    Yoshida, Hiroyuki; Nagayoshi, Takuji; Takase, Kazuyuki; Akimoto, Hajime

    Thermal-hydraulic design of the current boiling water reactor (BWR) is performed by correlations with empirical results of actual-size tests. However, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test of an embodiment of its design is required to confirm or modify such correlations. Development of a method that enables the thermal-hydraulic design of nuclear reactors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason we developed an advanced thermal-hydraulic design method for FLWRs using innovative two-phase flow simulation technology. In this study, detailed Two-Phase Flow simulation code using advanced Interface Tracking method: TPFIT is developed to calculate the detailed information of the two-phase flow. We tried to verify the TPFIT code by comparing it with the 2-channel air-water and steam-water mixing experimental results. The predicted result agrees well the observed results and bubble dynamics through the gap and cross flow behavior could be effectively predicted by the TPFIT code, and pressure difference between fluid channels is responsible for the fluid mixing.

  15. Thermal hydraulics development for CASL

    Energy Technology Data Exchange (ETDEWEB)

    Lowrie, Robert B [Los Alamos National Laboratory

    2010-12-07

    This talk will describe the technical direction of the Thermal-Hydraulics (T-H) Project within the Consortium for Advanced Simulation of Light Water Reactors (CASL) Department of Energy Innovation Hub. CASL is focused on developing a 'virtual reactor', that will simulate the physical processes that occur within a light-water reactor. These simulations will address several challenge problems, defined by laboratory, university, and industrial partners that make up CASL. CASL's T-H efforts are encompassed in two sub-projects: (1) Computational Fluid Dynamics (CFD), (2) Interface Treatment Methods (ITM). The CFD subproject will develop non-proprietary, scalable, verified and validated macroscale CFD simulation tools. These tools typically require closures for their turbulence and boiling models, which will be provided by the ITM sub-project, via experiments and microscale (such as DNS) simulation results. The near-term milestones and longer term plans of these two sub-projects will be discussed.

  16. GCFR thermal-hydraulic experiments

    International Nuclear Information System (INIS)

    The thermal-hydraulic experimental studies performed and planned for the Gas-Cooled Fast Reactor (GCFR) core assemblies are described. The experiments consist of basic studies performed to obtain correlations, and bundle experiments which provide input for code validation and design verification. These studies have been performed and are planned at European laboratories, US national laboratories, Universities in the US, and at General Atomic Company

  17. Benchmarking of thermal hydraulic loop models for Lead-Alloy Cooled Advanced Nuclear Energy System (LACANES), phase-I: Isothermal steady state forced convection

    International Nuclear Information System (INIS)

    As highly promising coolant for new generation nuclear reactors, liquid Lead-Bismuth Eutectic has been extensively worldwide investigated. With high expectation about this advanced coolant, a multi-national systematic study on LBE was proposed in 2007, which covers benchmarking of thermal hydraulic prediction models for Lead-Alloy Cooled Advanced Nuclear Energy System (LACANES). This international collaboration has been organized by OECD/NEA, and nine organizations - ENEA, ERSE, GIDROPRESS, IAEA, IPPE, KIT/IKET, KIT/INR, NUTRECK, and RRC KI - contribute their efforts to LACANES benchmarking. To produce experimental data for LACANES benchmarking, thermal-hydraulic tests were conducted by using a 12-m tall LBE integral test facility, named as Heavy Eutectic liquid metal loop for integral test of Operability and Safety of PEACER (HELIOS) which has been constructed in 2005 at the Seoul National University in the Republic of Korea. LACANES benchmark campaigns consist of a forced convection (phase-I) and a natural circulation (phase-II). In the forced convection case, the predictions of pressure losses based on handbook correlations and that obtained by Computational Fluid Dynamics code simulation were compared with the measured data for various components of the HELIOS test facility. Based on comparative analyses of the predictions and the measured data, recommendations for the prediction methods of a pressure loss in LACANES were obtained. In this paper, results for the forced convection case (phase-I) of LACANES benchmarking are described.

  18. Helical coil thermal hydraulic model

    International Nuclear Information System (INIS)

    A model has been developed in Matlab environment for the thermal hydraulic analysis of helical coil and shell steam generators. The model considers the internal flow inside one helix and its associated control volume of water on the external side, both characterized by their inlet thermodynamic conditions and the characteristic geometry data. The model evaluates the behaviour of the thermal-hydraulic parameters of the two fluids, such as temperature, pressure, heat transfer coefficients, flow quality, void fraction and heat flux. The evaluation of the heat transfer coefficients as well as the pressure drops has been performed by means of the most validated literature correlations. The model has been applied to one of the steam generators of the IRIS modular reactor and a comparison has been performed with the RELAP5/Mod.3.3 code applied to an inclined straight pipe that has the same length and the same elevation change between inlet and outlet of the real helix. The predictions of the developed model and RELAP5/Mod.3.3 code are in fairly good agreement before the dryout region, while the dryout front inside the helical pipes is predicted at a lower distance from inlet by the model

  19. Reliable prediction of complex thermal hydraulic parameters by ANN

    International Nuclear Information System (INIS)

    Thermal hydraulic data-base is very useful in the design and analysis of the proposed Advanced Heavy Water Reactor which relies on natural circulation for normal core cooling. Compilation of the thermal hydraulic data-base is in progress. Artificial Neural Networks (ANNs), have been applied to analyse the consistency and accuracy of the data-base. The ANN predictions are more accurate and cover wider range of parameters compared to model based predictions

  20. Problems of the thermal-hydraulic computer codes perfection

    International Nuclear Information System (INIS)

    The analysis of the current state of research and development in the field of thermal-hydraulic computer codes. The experience of the creation of domestic and foreign versions of the most advanced versions of code improved estimate. Considerable attention is paid to the problems of calculation of the critical heat fluxes in the channels of nuclear reactors. Considered problematic issues to ensure the reliability of thermal-hydraulic steam-generating channels in a thermoacoustic oscillation

  1. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    Energy Technology Data Exchange (ETDEWEB)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition

  2. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward

    International Nuclear Information System (INIS)

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high

  3. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    Energy Technology Data Exchange (ETDEWEB)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition

  4. Thermal Hydraulic Tests for Reactor Core Safety

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S. K.; Baek, W. P.; Chun, S. Y. (and others)

    2007-06-15

    The main objectives of the present project are to resolve the current issues of reactor core thermal hydraulics, to develop an advanced measurement and analytical techniques, and to perform reactor core safety verification tests. 6x6 reflood experiments, various heat transfer experiments using Freon, and experiments on the spacer grids effects on the post-dryout are carried out using spacer grids developed in Korea in order to resolve the current issues of the reactor core thermal hydraulics. In order to develop a reflood heat transfer model, the detailed reflood phenomena are visualized and measured using round tube and 2x2 rod bundle. A detailed turbulent mixing phenomenon for subchannels is measured using advanced measurement techniques such as LDV and PIV. MARS and MATRA codes developed in Korea are assessed, verified and improved using the obtained experimental data. Finally, a systematic quality assurance program and experimental data generation system has been constructed in order to increase the reliability of the experimental data.

  5. Application of Advanced Thermal Hydraulic TRACG Model to Preserve Operating Margins in BWRs at Extended Power Up-rate Conditions

    International Nuclear Information System (INIS)

    GE has developed TRACG, a customized BWR version of the TRAC model, for application to BWR analyses. This model was initially applied to special BWR challenges and for benchmarking the official simplified thermal-hydraulic design models. However, in past years extensive additional model development, qualification and application studies have been completed. This development has followed the CSAU methodology, where extensive model evaluation and qualification have been performed to demonstrate the applicability of the model and to quantify the uncertainty in the model parameters as well as in plant parameters and initial conditions. This has then been combined with a statistically based application methodology following the CSAU approach to generate tolerance limits for the critical safety and design parameters. This effort has resulted in application processes that have been reviewed and approved by the US NRC to enable routine application of the TRACG model to the design and licensing analyses and utilize the improved operating margin to optimize the fuel cycle design. These applications have been supported by development of programs that construct specific plant and problem base-decks that utilize BWR plant characteristics and system databases to standardize and streamline the application to several plants. The application of the TRACG model in Transient and LOCA analyses has assisted in allowing similar power peaking at higher power density conditions for BWRs. Also, the application of the TRACG model in Stability analyses has assisted in preserving the setpoints of stability monitoring systems to avoid margin loss for high power density applications. TRACG is being used for analysis of ATWS events. It has been used to support the development of emergency procedure guidelines, and it is currently being used to demonstrate that the suppression pool temperature limits can be met for up-rated conditions. Finally, the application of the TRACG model in Faulted Load

  6. Statistically based uncertainty analysis for ranking of component importance in the thermal-hydraulic safety analysis of the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    The Analytic Hierarchy Process (AHP) has been used to help determine the importance of components and phenomena in thermal-hydraulic safety analyses of nuclear reactors. The AHP results are based, in part on expert opinion. Therefore, it is prudent to evaluate the uncertainty of the AHP ranks of importance. Prior applications have addressed uncertainty with experimental data comparisons and bounding sensitivity calculations. These methods work well when a sufficient experimental data base exists to justify the comparisons. However, in the case of limited or no experimental data the size of the uncertainty is normally made conservatively large. Accordingly, the author has taken another approach, that of performing a statistically based uncertainty analysis. The new work is based on prior evaluations of the importance of components and phenomena in the thermal-hydraulic safety analysis of the Advanced Neutron Source Reactor (ANSR), a new facility now in the design phase. The uncertainty during large break loss of coolant, and decay heat removal scenarios is estimated by assigning a probability distribution function (pdf) to the potential error in the initial expert estimates of pair-wise importance between the components. Using a Monte Carlo sampling technique, the error pdfs are propagated through the AHP software solutions to determine a pdf of uncertainty in the system wide importance of each component. To enhance the generality of the results, study of one other problem having different number of elements is reported, as are the effects of a larger assumed pdf error in the expert ranks. Validation of the Monte Carlo sample size and repeatability are also documented

  7. Reactor core calculations incorporating subassembly thermal hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Lynas, S.W. [Applied Modelling and Computation Group Imperial Coll. Centre for Environmental Technology Royal School of Mines Prince Consort Road London (United Kingdom); Jones, J.R.

    1997-12-31

    Three dimensional reactor physics calculations performed in parallel with subassembly thermal hydraulic analysis can be used to examine local reactivity effects and increase modelling accuracy. Coupling together codes for coarse mesh neutronics and subassembly thermal hydraulics aids fault studies (fuel clad integrity, safety margin indication etc) and the examination of the interaction between physics and thermal hydraulics during transient events such as LOCA, boron dilution and control rod ejection. Local heating of the coolant decreases reactivity and the fission power peaking factor. Doppler feedback is stronger in the hot region of the fuel, also reducing peak power and reactivity. These thermal hydraulic feedback effects can play an important role in decelerating power excursions and their representation is described in this paper. (author)

  8. Reactor core calculations incorporating subassembly thermal hydraulics

    International Nuclear Information System (INIS)

    Three dimensional reactor physics calculations performed in parallel with subassembly thermal hydraulic analysis can be used to examine local reactivity effects and increase modelling accuracy. Coupling together codes for coarse mesh neutronics and subassembly thermal hydraulics aids fault studies (fuel clad integrity, safety margin indication etc) and the examination of the interaction between physics and thermal hydraulics during transient events such as LOCA, boron dilution and control rod ejection. Local heating of the coolant decreases reactivity and the fission power peaking factor. Doppler feedback is stronger in the hot region of the fuel, also reducing peak power and reactivity. These thermal hydraulic feedback effects can play an important role in decelerating power excursions and their representation is described in this paper. (author)

  9. MAAP thermal-hydraulic qualification studies

    International Nuclear Information System (INIS)

    The MAAP Thermal-Hydraulic Qualification and Application Project has as its objective to identify those thermal-hydraulic phenomena modeled in MAAP which are important in predicting severe accident sequences and to qualify those models. This report discusses sensitivity studies performed with MAAP to determine the sensitivity of the code to important input and modeling parameters and comparison of MAAP predictions to test data and independent predictions by other computer programs

  10. Thermal hydraulics and mechanics core design programs

    International Nuclear Information System (INIS)

    The report documents the work performed within the Research and Development Task Thermal hydraulics and mechanics core design programs, funded by the German government. It contains the development of new codes, the extension of existing codes, the qualification and verification of codes and the development of a code library. The overall goal of this work was to adapt the system of thermal hydraulics and mechanics codes to the permanently growing requirements of the status of science and technology

  11. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8∼2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4∼2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure

  12. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

  13. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Liles, D.R.; Mahaffy, J.H.

    1986-07-01

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients.

  14. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients

  15. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform the tests for design, operation, and safety regulation of pressurized water reactors. In the first phase of this project (1997.8∼2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished. In the second phase (2002.4∼2005.2), an optimized design of the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) was established and the construction of the facility was almost completed. In the third phase (2005.3∼2007.2), the construction and commission tests of the ATLAS are to be completed and some first-phase tests are to be conducted

  16. Ocean thermal gradient hydraulic power plant.

    Science.gov (United States)

    Beck, E J

    1975-07-25

    Solar energy stored in the oceans may be used to generate power by exploiting ploiting thermal gradients. A proposed open-cycle system uses low-pressure steam to elevate vate water, which is then run through a hydraulic turbine to generate power. The device is analogous to an air lift pump. PMID:17813707

  17. Research on the improvement of nuclear safety -Thermal hydraulic tests for reactor safety system-

    International Nuclear Information System (INIS)

    The present research aims at the development of the thermal hydraulic verification test technology for the safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. In this research, test facilities simulating the primary coolant system and safety system are being constructed for the design verification tests of the existing and advanced nuclear power plant. 97 figs, 14 tabs, 65 refs. (Author)

  18. Research on the improvement of nuclear safety -Thermal hydraulic tests for reactor safety system-

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Moon Kee; Park, Choon Kyung; Yang, Sun Kyoo; Chun, Se Yung; Song, Chul Hwa; Jun, Hyung Kil; Jung, Heung Joon; Won, Soon Yun; Cho, Yung Roh; Min, Kyung Hoh; Jung, Jang Hwan; Jang, Suk Kyoo; Kim, Bok Deuk; Kim, Wooi Kyung; Huh, Jin; Kim, Sook Kwan; Moon, Sang Kee; Lee, Sang Il [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-06-01

    The present research aims at the development of the thermal hydraulic verification test technology for the safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. In this research, test facilities simulating the primary coolant system and safety system are being constructed for the design verification tests of the existing and advanced nuclear power plant. 97 figs, 14 tabs, 65 refs. (Author).

  19. Thermal Hydraulic Stability in a Coaxial Thermosyphon

    Institute of Scientific and Technical Information of China (English)

    YANG Jianhui; LU Wenqiang; LI Qing; LI Qiang; ZHOU Yuan

    2005-01-01

    The heat transfer and thermal hydraulic stability in a two-phase thermosyphon with coaxial riser and down-comer has been experimentally investigated and theoretically analyzed to facilitate its application in cold neutron source. The flow in a coaxial thermosyphon was studied experimentally for a variety of heating rates, transfer tube lengths, charge capacities, and area ratios. A numerical analysis of the hydraulic balance between the driving pressure head and the resistance loss has also been performed. The results show that the presented coaxial thermosyphon has dynamic performance advantages relative to natural circulation in a boiling water reactor.

  20. Evolution of developments and applications of advanced thermal-hydraulics and neutronic codes. Conclusions from Annapolis Workshop and Ankara Seminar, Objectives of the Present Workshop

    International Nuclear Information System (INIS)

    In the nuclear reactor safety area, during the last 30-40 years, thermal-hydraulics has been one of the key disciplines for simulation and analysis of transient and accident scenarios and also for the definition of preventive and mitigative measures in relation to these scenarios. A workshop was organised by OECD/NEA-CSNI at Annapolis (1996) where codes, physical models, numeric and new computer architecture were examined. In parallel a Specialist meeting on instrumentation in two phase flows was held in Santa Barbara beginning of 1997 in order to investigate new techniques for getting measurements of new physical parameters necessary for assessing the new physical models. Among the different applications of thermal-hydraulic codes, the use of Best Estimate methods in safety evaluation is certainly one of the major challenges for which the safety and economic issues are quite important. For these reasons OECD/NEA-CSNI organised a seminar in Ankara in 1998 entirely devoted to the use of Best Estimate methods in thermal-hydraulics analysis. This seminar allowed to get a better view of where we were in such applications and which were the remaining problems and issues. The present workshop held in Barcelona beginning of year 2000 will be a good opportunity for providing an updated review of the gained progresses and for analysing if the objectives and programs are still progressing in the right direction. In order to do such exercise, we will first recall the questions which were raised in Annapolis and the main conclusions which were drawn from these questions. The conclusions of Ankara Meeting will be reviewed in a second step. Finally we will list the objectives of this workshop in Barcelona which is held in the continuity of Annapolis Workshop and Ankara Seminar. (authors)

  1. Thermal-hydraulics associated with nuclear education and research

    International Nuclear Information System (INIS)

    This article was the rerecording of the author's lecture at the fourth 'Future Energy Forum' (aiming at improving nuclear safety and economics) held in December 2010. The lecture focused on (1) importance of thermal hydraulics associated with nuclear education and research (critical heat flux, two-phase flow and multiphase flow), (2) emerging trend of maintenance engineering (fluid induced vibration, flow accelerated corrosion and stress corrosion cracks), (3) fostering sensible nuclear engineer with common engineering sense, (4) balanced curriculum of basics and advanced research, (5) computerized simulation and fluid mechanics, (6) crucial point of thermo hydraulics education (viscosity, flux, steam and power generation), (7) safety education and human resources development (indispensable technologies such as defence in depth) and (8) topics of thermo hydraulics research (vortices of curbed pipes and visualization of two-phase flow). (T. Tanaka)

  2. Thermal-hydraulics in BWR

    International Nuclear Information System (INIS)

    In the heat transferring flow in BWRs, the heightening of heat transfer performance accompanying the development of new fuel for the purpose of reducing spent fuel generation and the improvement of fuel economy, the heightening of performance and the reduction of size of various heat exchangers, the development of the safety devices, of which the constitution is simple, the reliability is high, and the operation is easy, and so on are expected. As for ABWRs, thermal output is 3926 MW, and electricity output is 1356 MW. The system constitution of ABWR is shown. The main change from BWR to ABWR is the adoption of internal pumps, reinforced concrete containment vessels and electric control rod drive. For evaluating the limit output of high burnup fuel assemblies, the subchannel analysis and the effect that spacers exert to the limit output are explained. The heat transferring flow in moisture separation heater, condenser and feed water heater is reported. The heat transferring flow in passive containment vessel cooling system of water wall type and condensing type is described. (K.I.)

  3. Thermal-hydraulic analysis of nuclear reactors

    CERN Document Server

    Zohuri, Bahman

    2015-01-01

    This text covers the fundamentals of thermodynamics required to understand electrical power generation systems and the application of these principles to nuclear reactor power plant systems. It is not a traditional general thermodynamics text, per se, but a practical thermodynamics volume intended to explain the fundamentals and apply them to the challenges facing actual nuclear power plants systems, where thermal hydraulics comes to play.  Written in a lucid, straight-forward style while retaining scientific rigor, the content is accessible to upper division undergraduate students and aimed at practicing engineers in nuclear power facilities and engineering scientists and technicians in industry, academic research groups, and national laboratories. The book is also a valuable resource for students and faculty in various engineering programs concerned with nuclear reactors. This book also: Provides extensive coverage of thermal hydraulics with thermodynamics in nuclear reactors, beginning with fundamental ...

  4. Thermal-hydraulics of actinide burner reactors

    International Nuclear Information System (INIS)

    As a part of conceptual study of actinide burner reactors, core thermal-hydraulic analyses were conducted for two types of reactor concepts, namely (1) sodium-cooled actinide alloy fuel reactor, and (2) helium-cooled particle-bed reactor, to examine the feasibility of high power-density cores for efficient transmutation of actinides within the maximum allowable temperature limits of fuel and cladding. In addition, calculations were made on cooling of actinide fuel assembly. (author)

  5. Computer analysis of thermal hydraulics for nuclear reactor safety

    International Nuclear Information System (INIS)

    This paper gives an overview of ANSTO's capability and recent research and development activities in thermal hydraulic modelling for nuclear reactor safety analysis, particularly for our research reactor, HIFAR (High Flux Australian Reactor) and its intended replacement, the Replacement Research Reactor (RRR). Several tools contribute to ANSTO's capability in thermal hydraulic modelling, including RELAP (developed in US) - a code for reactor system thermal-hydraulic analysis; CFS (developed in UK) - a general computational fluid dynamics code , which was used for thermal hydraulic analysis in reactor fuel elements; and HIZAPP (developed at ANSTO) - for coupling neutronics with thermal-hydraulics for reactor transient analysis

  6. Issues and future direction of thermal-hydraulics research and development in nuclear power reactors

    International Nuclear Information System (INIS)

    The paper archives the proceedings of an expert panel discussion on the issues and future direction of thermal-hydraulic research and development in nuclear power reactors held at the NURETH-14 conference in Toronto, Canada, in September 2011. Thermal-hydraulic issues related to both operating and advanced reactors are presented. Advances in thermal-hydraulics have significantly improved the performance of operating reactors. Further thermal-hydraulics research and development is continuing in both experimental and computational areas for operating reactors, reactors under construction or ready for near-term deployment, and advanced Generation-IV reactors. As the computing power increases, the fine-scale multi-physics computational models, coupled with the systems analysis code, are expected to provide answers to many challenging problems in both operating and advanced reactor designs

  7. Visualization of nuclear thermal-hydraulics

    International Nuclear Information System (INIS)

    The VISUALIZATION is the strong tool to understand the complicated two-phase flows, including boiling. Many problems on the Nuclear Thermal-hydraulics are multi-disciplinary problems, e.g., Interaction between Material and Fluid, Interaction between Radiation and Heat Transfer, and so on. For these complex interdisciplinary problems, visualization studies help to investigate the mechanisms. In this decade, the technology of the high-speed camera has been progressed drastically. One Mega-pixel image can be captured with a sampling rate of 7000 Hz upto several seconds continuously. The complicated two-phase phenomenon can be clearly visualized using these tools. The boiling phenomenon under the irradiation condition has different characteristics to that under the non-irradiation conditions. The phenomenon was called as the RISA, Radiation Induced Surface Activation. With the gamma-ray irradiation to oxidized material surface, many interesting phenomena have been reported including, boiling enhancement, corrosion protection, and so on. The visualization study with high-speed camera can reveal the mechanism of the nucleate boiling degradations. The RISA is the interdisciplinary problems, i.e., Interface area of Radiation, Material and Thermal-hydraulics. Other applications of the high-speed camera on the interdisciplinary thermal-hydraulic problems will be discussed. Also, the future researches using the visualization technology will be reviewed. (author)

  8. Development of numerical procedure for thermal hydraulic design of nuclear reactors with advanced two-fluid model (1). Improvement of numerical stability of advanced two-fluid model

    International Nuclear Information System (INIS)

    Two-fluid model is still useful to simulate two-phase flow in large domain such as rod bundles. However, two-fluid model include a lot of constitutive equations, and the two-fluid model has problems that the results of analyses depend on accuracy of constitutive equations. To solve these problems, we have been developing an advanced two-fluid model. In this model, an interface tracking method is combined with the two-fluid model to predict large interface structure behavior without any constitutive equations, and constitutive equations to evaluate the effects of small bubbles or droplets are only required. In this study, we modified the advanced two-fluid model to improve the stability of the numerical simulation and reduce the computational time. In this paper, we describe the modification performed in this study and the numerical results of two-phase flow in various flow conditions are shown. (author)

  9. Thermal hydraulic feasibility assessment for the Spent Nuclear Fuel Project

    International Nuclear Information System (INIS)

    A series of scoping analyses have been completed investigating the thermal-hydraulic performance and feasibility of the Spent Nuclear Fuel Project (SNFP) Integrated Process Strategy (IPS). The SNFP was established to develop engineered solutions for the expedited removal, stabilization, and storage of spent nuclear fuel from the K Basins at the U.S. Department of Energy's Hanford Site in Richland, Washington. The subject efforts focused on independently investigating, quantifying, and establishing the governing heat production and removal mechanisms for each of the IPS operations and configurations, obtaining preliminary results for comparison with and verification of other analyses, and providing technology-based recommendations for consideration and incorporation into the design bases for the SNFP. The goal was to develop a series fo thermal-hydraulic models that could respond to all process and safety-related issues that may arise pertaining to the SNFP. A series of sensitivity analyses were also performed to help identify those parameters that have the greatest impact on energy transfer and hence, temperature control. It is anticipated that the subject thermal-hydraulic models will form the basis for a series of advanced and more detailed models that will more accurately reflect the thermal performance of the IPS and alleviate the necessity for some of the more conservative assumptions and oversimplifications, as well as form the basis for the final process and safety analyses

  10. Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 1

    International Nuclear Information System (INIS)

    More than 100 papers were presented. The meeting was divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling

  11. Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 2

    International Nuclear Information System (INIS)

    More than 100 papers presented at the meeting were divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling

  12. Review of computational thermal-hydraulic modeling

    International Nuclear Information System (INIS)

    Corrosion of heat transfer tubing in nuclear steam generators has been a persistent problem in the power generation industry, assuming many different forms over the years depending on chemistry and operating conditions. Whatever the corrosion mechanism, a fundamental understanding of the process is essential to establish effective management strategies. To gain this fundamental understanding requires an integrated investigative approach that merges technology from many diverse scientific disciplines. An important aspect of an integrated approach is characterization of the corrosive environment at high temperature. This begins with a thorough understanding of local thermal-hydraulic conditions, since they affect deposit formation, chemical concentration, and ultimately corrosion. Computational Fluid Dynamics (CFD) can and should play an important role in characterizing the thermal-hydraulic environment and in predicting the consequences of that environment,. The evolution of CFD technology now allows accurate calculation of steam generator thermal-hydraulic conditions and the resulting sludge deposit profiles. Similar calculations are also possible for model boilers, so that tests can be designed to be prototypic of the heat exchanger environment they are supposed to simulate. This paper illustrates the utility of CFD technology by way of examples in each of these two areas. This technology can be further extended to produce more detailed local calculations of the chemical environment in support plate crevices, beneath thick deposits on tubes, and deep in tubesheet sludge piles. Knowledge of this local chemical environment will provide the foundation for development of mechanistic corrosion models, which can be used to optimize inspection and cleaning schedules and focus the search for a viable fix

  13. Progress of thermal-hydraulic research on SCWR in NPIC

    International Nuclear Information System (INIS)

    For advantages of higher efficiency of thermal cycle, simpler primary coolant circuit etc, Supercritical Water Reactor (SCWR) is much competitive in the Generation-IV nuclear system. Based on the supercritical water reactor engineering, NPIC has started related numerical analysis, test technology research, test facility design and some experimental researches of thermal hydraulics of supercritical water since 2005. The progress of thermal hydraulic research on SCW in NPIC in the past five years, will be introduced briefly, included numerical research of SCW thermal hydraulic performance, test technology research, design and construction of thermal hydraulic test facilities, and related experimental research of SCW flow resistance and heat transfer behavior. (author)

  14. Views on the future of thermal hydraulic modeling

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, M. [Purdue Univ., West Lafayette, IN (United States)

    1997-07-01

    It is essential for the U.S. NRC to sustain the highest level of the thermal-hydraulics and reactor safety research expertise and continuously improve their accident analysis capability. Such expertise should span over four different areas which are strongly related to each other. These are: (1) Reactor Safety Code Development, (2) Two-phase Flow Modeling, (3) Instrumentation and Fundamental Experimental Research, and (4) Separate Effect and Integral Test. The NRC is already considering a new effort in the area of advanced thermal-hydraulics effort. Its success largely depends on the availability of a significantly improved two-phase flow formulation and constitutive relations supported by detailed experimental data. Therefore, it is recommended that the NRC start significant research efforts in the areas of two-phase flow modeling, instrumentation, basic and separate effect experiments which should be pursued systematically and with clearly defined objectives. It is desirable that some international program is developed in this area. This paper is concentrated on those items in the thermal-hydraulic area which eventually determine the quality of future accident analysis codes.

  15. Assessment of applicability of TRAC-BF1 for thermal hydraulic instability

    International Nuclear Information System (INIS)

    To assess the stability of advanced light water reactors or steam generators in FBR, JAEA has been developing a prediction method for thermal-hydraulic instability based on system analysis code TRAC-BF1. In the present paper, thermal-hydraulic instability experiments were analyzed with TRAC-BF1 code and the applicability of the code for thermal-hydraulic instability was estimated. The heat flux of the boundary between stable and unstable region is in agreement with the experimental data within about 10% of error. The onset of instability can be predicted with TRAC-BF1. (author)

  16. TRIGA mixed HEU - LEU thermal hydraulic analysis

    International Nuclear Information System (INIS)

    It is generally known that now TRIGA SSR has a mixed HEU-LEU core. In order to increase the reactor fuel utilization by slightly rising the excess reactivity, one idea was to extract the central fuel pin from the fuel cluster. The paper is dedicated to the core safety thermal hydraulic analysis of the modified cluster configuration. Thermal hydraulic analysis was done by neutronic computation with 3DDT (three-dimensional diffusion) computer code. The results of these computations, regarding pin-power factors, were input for the thermal hydraulic analysis. We use for our analysis COBRA IV computer code. The configuration analyzed was of a TRIGA fuel cluster of 25 fuel elements in 5 x 5 square configuration. It was considered multi-channel geometry specific for the COBRA computations. For the reason of analysis two cooling modes were chosen: - two pumps with 660 l/sec flow rate; - one cooling pump with 330 l/s. The computations were done for 14 MW rating power. In both cooling modes the heat transfer is produced by fluid forced convection. The gap heat transfer coefficient used is 1.36 +4 W/m2K, which corresponds for a gap width of 1.27-3 cm according to the Safety Report. It is to emphasize that both HEU and LEU fuel types share the same thermal characteristics. The important safety characteristics for both fuels is maximum central temperature, which is limited to 750 deg. C. The reason for this limitation is the fuel composition. Above this temperature the U-ZrH alloy changes the phase and hydrogen is generated what increases the stresses into the fuel. Working with COBRA computer code and with power peaking factors resulting from neutronic analysis, the effect of central pin removal is given by comparing the temperatures of the fuel elements in normal and modified cluster. A small increase of about 1% is observed in latter case as compared with the normal one. At the same time the temperatures at 4 pins in the vicinity of the central water-filed space is decreased

  17. Single- and two-phase flow modeling for coupled neutronics / thermal-hydraulics transient analysis of advanced sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Nuclear power is nowadays in the front rank as regards helping to meet the growing worldwide energy demand while avoiding an excessive increase in greenhouse gas emissions. However, the operating nuclear power plants are mainly thermal-neutron reactors and, as such, can not be maintained on the basis of the currently identified uranium resources beyond one century at the present consumption rate. Sustainability of nuclear power thus involves closure of the fuel cycle through breeding. With a uranium-based fuel, breeding can only be achieved using a fast-neutron reactor. Sodium-cooled fast reactor (SFR) technology benefits from 400 reactor-years of accumulated experience and is thus a prime candidate for the implementation of so-called Generation-IV nuclear energy systems. In this context, the safety demonstration of SFRs remains a major Research and Development related issue. The current research aims at the development of a computational tool for the in-depth understanding of SFR core behaviour during accidental transients, particularly those including boiling of the coolant. An accurate modelling of the core physics during such transients requires the coupling between 3D neutron kinetics and thermal-hydraulics in the core, to account for the strong interactions between the two-phase coolant flow and power variations caused by the sodium void effect. The present study is specifically focused upon models for the representation of sodium two-phase flow. The extension of the thermal-hydraulics TRACE code, previously limited to the simulation of single-phase sodium flow, has been carried out through the implementation of equations-of-state and closure relations specific to sodium. The different correlations have then been implemented as options. From the validation study carried out, it has been possible to recommend a set of models which provide satisfactory results, while considering annular flow as the dominant regime up to dryout and a smooth breakdown of the

  18. Advanced thermal management materials

    CERN Document Server

    Jiang, Guosheng; Kuang, Ken

    2012-01-01

    ""Advanced Thermal Management Materials"" provides a comprehensive and hands-on treatise on the importance of thermal packaging in high performance systems. These systems, ranging from active electronically-scanned radar arrays to web servers, require components that can dissipate heat efficiently. This requires materials capable of dissipating heat and maintaining compatibility with the packaging and dye. Its coverage includes all aspects of thermal management materials, both traditional and non-traditional, with an emphasis on metal based materials. An in-depth discussion of properties and m

  19. LMR thermal hydraulics calculations in the US

    International Nuclear Information System (INIS)

    A wide range of thermal hydraulics computer codes have been developed by various organizations in the US. These codes cover an extensive range of purposes from within-assembly-wise pin temperature calculations to plant wide transient analysis. The codes are used for static analysis, for analysis of protected anticipated transients, and for analysis of a wide range of unprotected transients for the more recent inherently safe LMR designs. Some of these codes are plant-specific codes with properties of a specific plant built into them. Other codes are more general and can be applied to a number of plants or designs. These codes, and the purposes for which they have been used, are described

  20. Multiphase flow dynamics 5 nuclear thermal hydraulics

    CERN Document Server

    Kolev, Nikolay Ivanov

    2015-01-01

    This Volume 5 of the successful book package "Multiphase Flow Dynamics" is devoted to nuclear thermal hydraulics which is a substantial part of nuclear reactor safety. It provides knowledge and mathematical tools for adequate description of the process of transferring the fission heat released in materials due to nuclear reactions into its environment. It step by step introduces into the heat release inside the fuel, temperature fields in the fuels, the "simple" boiling flow in a pipe described using ideas of different complexity like equilibrium, non equilibrium, homogeneity, non homogeneity. Then the "simple" three-fluid boiling flow in a pipe is described by gradually involving the mechanisms like entrainment and deposition, dynamic fragmentation, collisions, coalescence, turbulence. All heat transfer mechanisms are introduced gradually discussing their uncertainty. Different techniques are introduced like boundary layer treatments or integral methods. Comparisons with experimental data at each step demons...

  1. Thermal hydraulic reactor safety analyses and experiments

    International Nuclear Information System (INIS)

    The report introduces the results of the thermal hydraulic reactor safety research performed in the Nuclear Engineering Laboratory of the Technical Research Centre of Finland (VTT) during the years 1972-1987. Also practical applications i.e. analyses for the safety authorities and power companies are presented. The emphasis is on description of the state-of-the-art know how. The report describes VTT's most important computer codes, both those of foreign origin and those developed at VTT, and their assessment work, VTT's own experimental research, as well as international experimental projects and other forms of cooperation VTT has participated in. Appendix 8 contains a comprehensive list of the most important publications and technical reports produced. They present the content and results of the research in detail.(orig.)

  2. THE THREE DIMENSIONAL THERMAL HYDRAULIC CODE BAGIRA.

    Energy Technology Data Exchange (ETDEWEB)

    KALINICHENKO,S.D.; KOHUT,P.; KROSHILIN,A.E.; KROSHILIN,V.E.; SMIRNOV,A.V.

    2003-05-04

    BAGIRA - a thermal-hydraulic program complex was primarily developed for using it in nuclear power plant simulator models, but is also used as a best-estimate analytical tool for modeling two-phase mixture flows. The code models allow consideration of phase transients and the treatment of the hydrodynamic behavior of boiling and pressurized water reactor circuits. It provides the capability to explicitly model three-dimensional flow regimes in various regions of the primary and secondary circuits such as, the mixing regions, circular downcomer, pressurizer, reactor core, main primary loops, the steam generators, the separator-reheaters. In addition, it is coupled to a severe-accident module allowing the analysis of core degradation and fuel damage behavior. Section II will present the theoretical basis for development and selected results are presented in Section III. The primary use for the code complex is to realistically model reactor core behavior in power plant simulators providing enhanced training tools for plant operators.

  3. Thermal hydraulic design of intermediate heat exchanger

    International Nuclear Information System (INIS)

    Intermediate heat exchanger (IHX) is a very important component of Fast Breeder Reactor because it forms the boundary between radioactive primary sodium and non-radioactive secondary sodium. IHX of the 500 MWe Prototype Fast Breeder Reactor is a shell and tube heat exchanger with primary sodium flowing on the shell side. Cross flow heat transfer at the primary sodium entrance demands unequal secondary flow distribution in various tubes to ensure good safety margin in structural design. This paper brings out details of thermal hydraulic studies to arrive at a suitable secondary flow distribution and choice of a suitable flow distribution device to achieve the same. Application of two-dimensional analysis with computer code THYC-2D has been brought out. (author). 5 refs., 14 figs., 2 tabs

  4. Multiphase Flow Dynamics 5 Nuclear Thermal Hydraulics

    CERN Document Server

    Kolev, Nikolay Ivanov

    2012-01-01

    The present Volume 5 of the successful book package "Multiphase Flow Dynamics" is devoted to nuclear thermal hydraulics which is a substantial part of nuclear reactor safety. It provides knowledge and mathematical tools for adequate description of the process of transferring the fission heat released in materials due to nuclear reactions into its environment. It step by step introduces into the heat release inside the fuel, temperature fields in the fuels, the "simple" boiling flow in a pipe described using ideas of different complexity like equilibrium, non equilibrium, homogeneity, non homogeneity. Then the "simple" three-fluid boiling flow in a pipe is described by gradually involving the mechanisms like entrainment and deposition, dynamic fragmentation, collisions, coalescence, turbulence. All heat transfer mechanisms are introduced gradually discussing their uncertainty. Different techniques are introduced like boundary layer treatments or integral methods. Comparisons with experimental data at each step...

  5. First results of the ThAI thermal-hydraulic containment tests

    International Nuclear Information System (INIS)

    The new-built ThAI test facility is designed to fulfill those validation requirements in the areas of thermal hydraulics and fission product behaviour, in particular iodine. More precise data from thermal-hydraulic experiments are needed for validation of lumped-parameter codes simulating severe accident sequences, e.g. for the containment code system COCOSYS presently under development at GRS. Furthermore, advanced measurement techniques are applied in ThAI to comply with the requirements of CFD codes for detailed data of, e.g., convection flow fields. Concerning fission products, ThAI aims at investigating mass transport phenomena of volatile iodine at a technical scale. This is necessary because iodine mass transport modelling is so far based on small-scale experiments, which cannot reproduce effects of real thermal-hydraulic conditions in severe accidents such as free natural convection flows or stratification of sump and atmosphere. For ThAI, radioactive iodine 123I is used as a tracer to allow accurate measurements of iodine at low, accident-relevant concentrations. The ThAI test programme consists of a thermal-hydraulic part starting end 2000, and an iodine part to be performed in 2001/2002. Code calculations for the first block of thermal-hydraulic experiments have been made well in advance to use the unique opportunity of predicting the thermal hydraulics of a still unknown facility and thus demonstrate the state of the art of the codes and their application. (author)

  6. Thermal hydraulic studies of high temperature reactors

    International Nuclear Information System (INIS)

    The development of High Temperature Nuclear Reactors capable of supplying process heat at a temperature around 1273 K, is in Progress at BARC. These nuclear reactors are being developed with the objective of providing energy to facilitate combined production of hydrogen, electricity, and drinking water. The reject and waste heat in the overall energy scheme are utilised for electricity generation and desalination, respectively. Presently, technology development for a small power (100 kWth) Compact High Temperature Reactor (CHTR) capable of supplying high temperature process heat at 1273 K is being carried out. In addition conceptual details of a 10 MWth reactor supplying heat at 1273 K for commercial hydrogen production, are also being worked out. 3D CFD analysis of the CHTR reactor core has been carried out to estimate the core heat removal capability by natural circulation during normal operating conditions. PHOENICS, a generalized CFD code is used for the analysis. The full-scale core, including fuel tube, coolant channel, plenums, down comer, heat sink, moderator and reflector has been modeled and analysed in PHOENICS. Steady state analysis is carried out to find flow distribution in the coolant circuit and temperature distribution in the whole core. Analyses have also been carried out to simulate various operational transients and accidental conditions of the reactor. This paper deals with the detailed CFD analysis. The details on the selection of the appropriate turbulence model, turbulent Prandtl number and mesh distribution for the CFD analysis are described in the paper. The results of the steady state and transient analyses are also presented in the paper. Paper shows one of the results of 3D CFD analysis for CHTR core. This paper also deals with the core thermal hydraulic analysis of the conceptual design of the 10MWth High Temperature Pebble Bed Reactor. Preliminary thermal hydraulic analysis is carried out with FLiBe as the primary coolants. The

  7. Thermal hydraulic analysis of nuclear research reactors

    International Nuclear Information System (INIS)

    A loss of coolant accident (LOCA) can cause total or partial core uncovery which is followed by substantial fuel element temperature increase due to fuel residual heat. It is essential to demonstrate that such a temperature increase does not lead to excessive core melting and to significant radioactive material release into the reactor building and consequently to the environment. The THEAP computer codes able to perform reliable analysis of such accidents have been developed. THEAP-I is a computer code developed with the aim to contribute to the safety analysis of the MTR open pool research reactors. THEAP-I is designed for three dimensional, transient thermal/hydraulic analysis of a thermally interacting channel bundle totally immersed into water or air, such as the reactor core. The mathematical and physical models and methods of the solution are given as well as the code description and the input data. A sample problem is included, referring to the Greek Research Reactor analysis, under a hypothetical severe loss of coolant accident. The micro computer version of the code is also described. More emphasis is given in the new features of the code (i.e. input data structure). A set of instructions for running in an IBM-AT2 computer with the microsoft FORTRAN V4.0 is included together with a sample problem referring to the Greek Research Reactor. THEAP-I can be used also for other MTR open pool research reactors. Refs and figs

  8. Visual and intelligent transients and accidents analyzer based on thermal-hydraulic system code

    International Nuclear Information System (INIS)

    Full text of publication follows: Many thermal-hydraulic system codes were developed in the past twenty years, such as RELAP5, RETRAN, ATHLET, etc. Because of their general and advanced features in thermal-hydraulic computation, they are widely used in the world to analyze transients and accidents. But there are following disadvantages for most of these original thermal-hydraulic system codes. Firstly, because models are built through input decks, so the input files are complex and non-figurative, and the style of input decks is various for different users and models. Secondly, results are shown in off-line data file form. It is not convenient for analysts who may pay more attention to dynamic parameters trend and changing. Thirdly, there are few interfaces with other program in these original thermal-hydraulic system codes. This restricts the codes expanding. The subject of this paper is to develop a powerful analyzer based on these thermal-hydraulic system codes to analyze transients and accidents more simply, accurately and fleetly. Firstly, modeling is visual and intelligent. Users build the thermalhydraulic system model using component objects according to their needs, and it is not necessary for them to face bald input decks. The style of input decks created automatically by the analyzer is unified and can be accepted easily by other people. Secondly, parameters concerned by analyst can be dynamically communicated to show or even change. Thirdly, the analyzer provide interface with other programs for the thermal-hydraulic system code. Thus parallel computation between thermal-hydraulic system code and other programs become possible. In conclusion, through visual and intelligent method, the analyzer based on general and advanced thermal-hydraulic system codes can be used to analysis transients and accidents more effectively. The main purpose of this paper is to present developmental activities, assessment and application results of the visual and intelligent

  9. INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

    Energy Technology Data Exchange (ETDEWEB)

    Glenn McCreery; Hugh McIlroy

    2007-09-01

    Advanced computer modeling and simulation tools and protocols will be heavily relied on for a wide variety of system studies, engineering design activities, and other aspects of the Next Generation Nuclear Power (NGNP) Very High Temperature Reactor (VHTR), the DOE Global Nuclear Energy Partnership (GNEP), and light-water reactors. The goal is for all modeling and simulation tools to be demonstrated accurate and reliable through a formal Verification and Validation (V&V) process, especially where such tools are to be used to establish safety margins and support regulatory compliance, or to design a system in a manner that reduces the role of expensive mockups and prototypes. Recent literature identifies specific experimental principles that must be followed in order to insure that experimental data meet the standards required for a “benchmark” database. Even for well conducted experiments, missing experimental details, such as geometrical definition, data reduction procedures, and manufacturing tolerances have led to poor Benchmark calculations. The INL has a long and deep history of research in thermal hydraulics, especially in the 1960s through 1980s when many programs such as LOFT and Semiscle were devoted to light-water reactor safety research, the EBRII fast reactor was in operation, and a strong geothermal energy program was established. The past can serve as a partial guide for reinvigorating thermal hydraulic research at the laboratory. However, new research programs need to fully incorporate modern experimental methods such as measurement techniques using the latest instrumentation, computerized data reduction, and scaling methodology. The path forward for establishing experimental research for code model validation will require benchmark experiments conducted in suitable facilities located at the INL. This document describes thermal hydraulic facility requirements and candidate buildings and presents examples of suitable validation experiments related

  10. INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

    International Nuclear Information System (INIS)

    Advanced computer modeling and simulation tools and protocols will be heavily relied on for a wide variety of system studies, engineering design activities, and other aspects of the Next Generation Nuclear Power (NGNP) Very High Temperature Reactor (VHTR), the DOE Global Nuclear Energy Partnership (GNEP), and light-water reactors. The goal is for all modeling and simulation tools to be demonstrated accurate and reliable through a formal Verification and Validation (V and V) process, especially where such tools are to be used to establish safety margins and support regulatory compliance, or to design a system in a manner that reduces the role of expensive mockups and prototypes. Recent literature identifies specific experimental principles that must be followed in order to insure that experimental data meet the standards required for a 'benchmark' database. Even for well conducted experiments, missing experimental details, such as geometrical definition, data reduction procedures, and manufacturing tolerances have led to poor Benchmark calculations. The INL has a long and deep history of research in thermal hydraulics, especially in the 1960s through 1980s when many programs such as LOFT and Semiscle were devoted to light-water reactor safety research, the EBRII fast reactor was in operation, and a strong geothermal energy program was established. The past can serve as a partial guide for reinvigorating thermal hydraulic research at the laboratory. However, new research programs need to fully incorporate modern experimental methods such as measurement techniques using the latest instrumentation, computerized data reduction, and scaling methodology. The path forward for establishing experimental research for code model validation will require benchmark experiments conducted in suitable facilities located at the INL. This document describes thermal hydraulic facility requirements and candidate buildings and presents examples of suitable validation experiments related

  11. Thermal-hydraulic modeling needs for passive reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, J.M. [Nuclear Regulatory Commission, Washington, DC (United States)

    1997-07-01

    The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered, but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken.

  12. Development and validation of thermal hydraulic code in rolling motion

    International Nuclear Information System (INIS)

    The RELAP5/MOD3.3 code was modified by adding a module calculating the effect of rolling motion and introducing new flow and heat transfer models. The thermal hydraulic code in rolling motion was developed. The experimental data were used to validate the theoretical models and calculation results. It is shown that the new flow and heat transfer models can correctly calculate the frictional resistance and heat transfer coefficients in rolling motion. The developed thermal hydraulic code can be used to simulate the thermal hydraulic system in rolling motion. (authors)

  13. Thermal-hydraulics of the Loviisa reactor pressure vessel overcooling transients

    International Nuclear Information System (INIS)

    In the Loviisa reactor pressure vessel safety analyses, the thermal-hydraulics of various overcooling transients has been evaluated to give pertinent initial data for fracture-mechanics calculations. The thermal-hydraulic simulations of the developed overcooling scenarios have been performed using best-estimate thermal-hydraulic computer codes. Experimental programs have been carried out to study phenomena related to natural circulation interruptions in the reactor coolant system. These experiments include buoyancy-induced phenomena such as thermal mixing and stratification of cold high-pressure safety injection water in the cold legs and the downcomer, and oscillations of the single-phase natural circulation. In the probabilistic pressurized thermal shock study, the Loviisa training simulator and the advanced system code RELAP5/MOD2 were utilized to simulate selected sequences. Flow stagnation cases were separately calculated with the REMIX computer program. The methods employed were assessed for these calculations against the plant data and own experiments

  14. Thermal-hydraulic modeling and analysis of hydraulic system by pseudo-bond graph

    Institute of Scientific and Technical Information of China (English)

    胡均平; 李科军

    2015-01-01

    To increase the efficiency and reliability of the thermodynamics analysis of the hydraulic system, the method based on pseudo-bond graph is introduced. According to the working mechanism of hydraulic components, they can be separated into two categories: capacitive components and resistive components. Then, the thermal-hydraulic pseudo-bond graphs of capacitive C element and resistance R element were developed, based on the conservation of mass and energy. Subsequently, the connection rule for the pseudo-bond graph elements and the method to construct the complete thermal-hydraulic system model were proposed. On the basis of heat transfer analysis of a typical hydraulic circuit containing a piston pump, the lumped parameter mathematical model of the system was given. The good agreement between the simulation results and experimental data demonstrates the validity of the modeling method.

  15. Advanced Hydraulic Studies on Enhancing Particle Removal

    DEFF Research Database (Denmark)

    He, Cheng

    The removal of suspended solids and attached pollutants is one of the main treatment processes in wastewater treatment. This thesis presents studies on the hydraulic conditions of various particle removal facilities for possible ways to increase their treatment capacity and performance by utilizing...... and improving hydraulic conditions. Unlike most traditional theses which usually focus only on one particular subject of study, this thesis contains four relatively independent studies which cover the following topics: a newly proposed particle settling enhancement plate, the redesign of the inlet zone...... of a high-flow rate clarifier, identify the hydraulic problems of an old partially functioned CSO facility and investigate possible ways to entirely eliminate untreated CSO by improving its hydraulic capacity and performance. In order to be easily understood, each part includes its own abstract...

  16. Thermal-hydraulic design of the 200 MW NHR

    International Nuclear Information System (INIS)

    The thermal hydraulic design of the 200-MW Nuclear Heating Reactor (NHR), design criteria, design methods, important characteristics and some development results are presented in this paper. (author). 5 refs, 8 figs, 2 tabs

  17. Thermal-hydraulics for space power, propulsion, and thermal management system design

    International Nuclear Information System (INIS)

    The present volume discusses thermal-hydraulic aspects of current space projects, Space Station thermal management systems, the thermal design of the Space Station Free-Flying Platforms, the SP-100 Space Reactor Power System, advanced multi-MW space nuclear power concepts, chemical and electric propulsion systems, and such aspects of the Space Station two-phase thermal management system as its mechanical pumped loop and its capillary pumped loop's supporting technology. Also discussed are the startup thaw concept for the SP-100 Space Reactor Power System, calculational methods and experimental data for microgravity conditions, an isothermal gas-liquid flow at reduced gravity, low-gravity flow boiling, computations of Space Shuttle high pressure cryogenic turbopump ball bearing two-phase coolant flow, and reduced-gravity condensation

  18. Validating and Verifying a New Thermal-Hydraulic Analysis Tool

    International Nuclear Information System (INIS)

    The Idaho National Engineering and Environmental Laboratory (INEEL) has developed a new analysis tool by coupling the Fluent computational fluid dynamics (CFD) code to the RELAP5-3DC/ATHENA advanced thermal-hydraulic analysis code. This tool enables researchers to perform detailed, three-dimensional analyses using Fluent's CFD capability while the boundary conditions required by the Fluent calculation are provided by the balance-of-system model created using RELAP5-3DC/ATHENA. Both steady-state and transient calculations can be performed, using many working fluids and point to three-dimensional neutronics. A general description of the techniques used to couple the codes is given. The validation and verification (V and V) matrix is outlined. V and V is presently ongoing. (authors)

  19. Thermal-hydraulic characteristic of the PGV-1000 steam generator

    International Nuclear Information System (INIS)

    Horizontal steam generators are typical parts of nuclear power plants with pressure water reactor type VVER. By means of this computer program, a detailed thermal-hydraulic study of the horizontal steam generator PGV-1000 has been carried out and a special attention has been paid to the thermal-hydraulics of the secondary side. A set of important steam generator characteristics has been obtained and analyzed. Some of the interesting results of the analysis are presented in the paper. (author)

  20. Thermal-hydraulic modeling of reactivity accidents in MTR reactors

    OpenAIRE

    Khater Hany; Abu-El-Maty Talal; El-Morshdy El-Din Salah

    2006-01-01

    This paper describes the development of a dynamic model for the thermal-hydraulic analysis of MTR research reactors during a reactivity insertion accident. The model is formulated for coupling reactor kinetics with feedback reactivity and reactor core thermal-hydraulics. To represent the reactor core, two types of channels are considered, average and hot channels. The developed computer program is compiled and executed on a personal computer, using the FORTRAN language. The model is validated...

  1. Development of visualization software for thermal-hydraulic analysis in a tight-lattice bundle using AVS

    International Nuclear Information System (INIS)

    Thermal-hydraulic safety in a tight-lattice bundle has been analyzed to contribute thermal design of an advanced water-cooled reactor core. Since the analytical geometry is complicated, it is difficult to understand the analysis results using general visualization software. In this study, the visualization program for the thermal-hydraulic analysis in the tight-lattice bundle was developed using the software AVS/Express. It can reproduce the three-dimensional view and graphs of the analysis results and it is helpful in understanding the thermal-hydraulic phenomena in the tight-lattice bundle. (author)

  2. Thermal-hydraulic analysis of a cylindrical blanket module using ATHENA code

    International Nuclear Information System (INIS)

    ATHENA (Advanced Thermal-Hydraulic Energy Network Analyzer) is a new computer code for thermal-hydraulic analyses of many energy systems. Multiple-loop and multiple-fluid capabilities have been emphasized during the code development. A pilot version of ATHENA has incorporated a fusion kinetic package to model the effect of first wall temperature variation on the reactor conditions. The capability has been demonstrated by analyzing the performance under various conditions of a cylindrical fusion blanket module. The results have shown the viability of using ATHENA for fusion reactor design and safety analyses

  3. Thermal Hydraulic Calculation Of PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Thermal hydraulic calculation for 1 MW (thermal) PUSPATI TRIGA Reactor will be carried out using COOLOD-N2. COOLOD-N2 was developed by Japan Atomic Energy Agency and this code has capability in calculating fuel temperature distribution, coolant temperature heat flux as well as departure nucleate boiling (DNB) heat flux. For the case of RTP, the parameter such as cladding and fuel meat temperature, inlet and outlet coolant temperature was calculated in order to obtain the heat flux and DNB ratio. This paper will discuss and compare the steady state thermal hydraulic calculation for RTP and some safety parameters stated in RTP Safety Analysis Report (SAR). (author)

  4. 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.

  5. Advanced Control Strategies for Mobile Hydraulic Applications

    OpenAIRE

    Cristofori, Davide

    2013-01-01

    Mobile hydraulic machines are affected by numerous undesired dynamics, mainly discontinuous motion and vibrations. Over the years, many methods have been developed to limit the extent of those undesired dynamics and improve controllability and safety of operation of the machine. However, in most of the cases, today's methods do not significantly differ from those developed in a time when electronic controllers were slower and less reliable than they are today. This dissertation addresses t...

  6. Thermal hydraulic test and verification of thermal hydraulic computer code for two-phase flow in U-bend tube (Part 2)

    International Nuclear Information System (INIS)

    In the U-bend region of steam generators of PWR plants, we have experienced some tube failures due to flow induced vibration. Therefore, it is important to confirm the reliability of U-bend tubes from the view point of flow induced vibration in order to improve the reliability of PWR plants. NUPEC (Nuclear Power Engineering Corporation) has initiated a new steam generator project sponsored by MITI (Ministry of International Trade and Industry) in 1993 fiscal year in order to verify the reliability of U-bend tubes of steam generators. Several tests on the thermal hydraulics and the flow induced vibration of the steam generators U-bend region have already been performed (1), (2) elsewhere in the world. However, all tests were performed using small-scale model steam generators. It is therefore expected that this research program will provide valuable data to advance the evaluation techniques of thermal hydraulic flow and flow induced vibration for steam generators. This paper focuses on measuring the void fraction and interfacial velocity profiles in U-bend tubes, and verifying the thermal hydraulic computer code FIT-OEL by using these data to upgrade the accuracy of prediction for actual SG thermal hydraulic behaviors. (author)

  7. Experimental validation of the thermal-hydraulic code SACATRI

    Energy Technology Data Exchange (ETDEWEB)

    Merroun, O., E-mail: meroun.ossama@gmail.co [LMR/ERSN, Department of Physics, Faculty of Sciences, Abdelmalek Essaadi University, B.P. 2121, Tetouan (Morocco); Al Mers, A. [Department of Energetics, Ecole Nationale Superieure d' Arts et Metiers, Moulay Ismail University, B.P. 4024, Meknes (Morocco); Veloso, M.A. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN), Belo Horizonte, MG (Brazil); El Bardouni, T.; El Bakkari, B. [LMR/ERSN, Department of Physics, Faculty of Sciences, Abdelmalek Essaadi University, B.P. 2121, Tetouan (Morocco); Chakir, E. [LRM/EPTN, Department of Physics, Faculty of Sciences, Kenitra (Morocco)

    2009-12-15

    A sub-channel analysis steady state thermal-hydraulic code (SACATRI) was developed for the Moroccan TRIGA MARK II research reactor. The main objective of the thermal-hydraulic study of the whole reactor core is to evaluate the main safety parameters of the reactor core, and to ensure that they are within the safety limits for any operating conditions. The thermal-hydraulic model used in SACATRI is based on four partial differential equations that describe the conservation of mass, energy and momentum. In order to assess the thermal-hydraulic mathematical model of SACATRI, the present paper focuses on the quantification of the physical model accuracy to judge if the code is capable to represent the thermal-hydraulic behaviour of the reactor core with sufficient accuracy. The methodology adopted is based on the comparison between responses from SACATRI computational model and experimentally measured responses performed on the IPR-R1 TRIGA research reactor. The results showed good agreement between SACATRI predictions and the experimental measurements where the discrepancies observed (simulation-experiment) are less than 6%.

  8. Experimental validation of the thermal-hydraulic code SACATRI

    International Nuclear Information System (INIS)

    A sub-channel analysis steady state thermal-hydraulic code (SACATRI) was developed for the Moroccan TRIGA MARK II research reactor. The main objective of the thermal-hydraulic study of the whole reactor core is to evaluate the main safety parameters of the reactor core, and to ensure that they are within the safety limits for any operating conditions. The thermal-hydraulic model used in SACATRI is based on four partial differential equations that describe the conservation of mass, energy and momentum. In order to assess the thermal-hydraulic mathematical model of SACATRI, the present paper focuses on the quantification of the physical model accuracy to judge if the code is capable to represent the thermal-hydraulic behaviour of the reactor core with sufficient accuracy. The methodology adopted is based on the comparison between responses from SACATRI computational model and experimentally measured responses performed on the IPR-R1 TRIGA research reactor. The results showed good agreement between SACATRI predictions and the experimental measurements where the discrepancies observed (simulation-experiment) are less than 6%.

  9. Proceedings of the OECD/CSNI workshop on transient thermal-hydraulic and neutronic codes requirements

    International Nuclear Information System (INIS)

    This is a report on the CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements held at Annapolis, Maryland, USA November 5-8, 1996. This experts' meeting consisted of 140 participants from 21 countries; 65 invited papers were presented. The meeting was divided into five areas: (1) current and prospective plans of thermal hydraulic codes development; (2) current and anticipated uses of thermal-hydraulic codes; (3) advances in modeling of thermal-hydraulic phenomena and associated additional experimental needs; (4) numerical methods in multi-phase flows; and (5) programming language, code architectures and user interfaces. The workshop consensus identified the following important action items to be addressed by the international community in order to maintain and improve the calculational capability: (a) preserve current code expertise and institutional memory, (b) preserve the ability to use the existing investment in plant transient analysis codes, (c) maintain essential experimental capabilities, (d) develop advanced measurement capabilities to support future code validation work, (e) integrate existing analytical capabilities so as to improve performance and reduce operating costs, (f) exploit the proven advances in code architecture, numerics, graphical user interfaces, and modularization in order to improve code performance and scrutibility, and (g) more effectively utilize user experience in modifying and improving the codes

  10. Proceedings of the OECD/CSNI workshop on transient thermal-hydraulic and neutronic codes requirements

    Energy Technology Data Exchange (ETDEWEB)

    Ebert, D.

    1997-07-01

    This is a report on the CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements held at Annapolis, Maryland, USA November 5-8, 1996. This experts` meeting consisted of 140 participants from 21 countries; 65 invited papers were presented. The meeting was divided into five areas: (1) current and prospective plans of thermal hydraulic codes development; (2) current and anticipated uses of thermal-hydraulic codes; (3) advances in modeling of thermal-hydraulic phenomena and associated additional experimental needs; (4) numerical methods in multi-phase flows; and (5) programming language, code architectures and user interfaces. The workshop consensus identified the following important action items to be addressed by the international community in order to maintain and improve the calculational capability: (a) preserve current code expertise and institutional memory, (b) preserve the ability to use the existing investment in plant transient analysis codes, (c) maintain essential experimental capabilities, (d) develop advanced measurement capabilities to support future code validation work, (e) integrate existing analytical capabilities so as to improve performance and reduce operating costs, (f) exploit the proven advances in code architecture, numerics, graphical user interfaces, and modularization in order to improve code performance and scrutibility, and (g) more effectively utilize user experience in modifying and improving the codes.

  11. Horizontal steam generator PGV-1000 thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    A computer program for the steady state thermal-hydraulic analysis of horizontal steam generator PGV-1000 is presented. The program provides the capability to analyze steam generator PGV-1000 primary side flow and temperature distribution, primary side pressure drops, heat transfer between the primary and secondary sides and multidimensional heat flux distribution. A special attention is paid to the thermal-hydraulics of the secondary side. The code predicts 3-D distribution of the void fraction at the secondary side, mass redistribution under the submerged perforated sheet and the steam generator level profile. By means of developed computer program a detailed thermal-hydraulic study of the PGV-1000 has been carried out. A wide range of calculations has been performed and a set of important steam generator characteristics has been obtained. Some of them are presented in the paper. (orig.)

  12. Horizontal steam generator PGV-1000 thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ubra, O. [Skoda Company, Prague (Switzerland); Doubek, M. [Czech Technical Univ., Prague (Switzerland)

    1995-12-31

    A computer program for the steady state thermal-hydraulic analysis of horizontal steam generator PGV-1000 is presented. The program provides the capability to analyze steam generator PGV-1000 primary side flow and temperature distribution, primary side pressure drops, heat transfer between the primary and secondary sides and multidimensional heat flux distribution. A special attention is paid to the thermal-hydraulics of the secondary side. The code predicts 3-D distribution of the void fraction at the secondary side, mass redistribution under the submerged perforated sheet and the steam generator level profile. By means of developed computer program a detailed thermal-hydraulic study of the PGV-1000 has been carried out. A wide range of calculations has been performed and a set of important steam generator characteristics has been obtained. Some of them are presented in the paper. (orig.). 5 refs.

  13. Current and anticipated uses of thermal hydraulic codes in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung-Doo; Chang, Won-Pyo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-07-01

    In Korea, the current uses of thermal hydraulic codes are categorized into 3 areas. The first application is in designing both nuclear fuel and NSSS. The codes have usually been introduced based on the technology transfer programs agreed between KAERI and the foreign vendors. Another area is in the supporting of the plant operations and licensing by the utility. The third category is research purposes. In this area assessments and some applications to the safety issue resolutions are major activities using the best estimate thermal hydraulic codes such as RELAP5/MOD3 and CATHARE2. Recently KEPCO plans to couple thermal hydraulic codes with a neutronics code for the design of the evolutionary type reactor by 2004. KAERI also plans to develop its own best estimate thermal hydraulic code, however, application range is different from KEPCO developing code. Considering these activities, it is anticipated that use of the best estimate hydraulic analysis code developed in Korea may be possible in the area of safety evaluation within 10 years.

  14. Current and anticipated uses of thermal hydraulic codes in Korea

    International Nuclear Information System (INIS)

    In Korea, the current uses of thermal hydraulic codes are categorized into 3 areas. The first application is in designing both nuclear fuel and NSSS. The codes have usually been introduced based on the technology transfer programs agreed between KAERI and the foreign vendors. Another area is in the supporting of the plant operations and licensing by the utility. The third category is research purposes. In this area assessments and some applications to the safety issue resolutions are major activities using the best estimate thermal hydraulic codes such as RELAP5/MOD3 and CATHARE2. Recently KEPCO plans to couple thermal hydraulic codes with a neutronics code for the design of the evolutionary type reactor by 2004. KAERI also plans to develop its own best estimate thermal hydraulic code, however, application range is different from KEPCO developing code. Considering these activities, it is anticipated that use of the best estimate hydraulic analysis code developed in Korea may be possible in the area of safety evaluation within 10 years

  15. Comparative Analysis of CTF and Trace Thermal-Hydraulic Codes Using OECD/NRC PSBT Benchmark Void Distribution Database

    OpenAIRE

    Avramova, M.; A. Velazquez-Lozada; Rubin, A.

    2013-01-01

    The international OECD/NRC PSBT benchmark has been established to provide a test bed for assessing the capabilities of thermal-hydraulic codes and to encourage advancement in the analysis of fluid flow in rod bundles. The benchmark was based on one of the most valuable databases identified for the thermal-hydraulics modeling developed by NUPEC, Japan. The database includes void fraction and departure from nucleate boiling measurements in a representative PWR fuel assembly. On behalf of the be...

  16. Thermal hydraulic calculations for TNRR using low enriched fuel

    International Nuclear Information System (INIS)

    This paper presents the preliminary results of the thermal-hydraulic calculations of the Tajoura nuclear research reactor (TNRR)) using low enrichment fuel 36%. The study considered the fresh core compact load. With the assumptions used in the calculation, no significant change in the thermal hydraulic parameters was noticed. A comparison was made for the most heated element between the (80% U235) enriched fuel and the (36% U235) enriched fuel under normal operating conditions. a slight increase in the maximum fuel surface temperature was noticed. These changes were due to change in the fuel material which is oxide fuel and to the increase in the meat thickness. (author)

  17. Large-scale simulations on thermal-hydraulics in fuel bundles of advanced nuclear reactors (Annual Report of the Earth Simulator Center, Dec 2008, 2007 issue)

    International Nuclear Information System (INIS)

    In order to predict the water-vapor two-phase flow dynamics in a fuel bundle of an advanced light-water reactor, large-scale numerical simulations were performed using a highly parallel-vector supercomputer, the earth simulator. Although conventional analysis methods such as subchannel codes and system analysis codes need composition equations based on the experimental data, it is difficult to obtain high prediction accuracy when experimental data to obtain the composition equations. Then, the present large-scale direct simulation method of water-vapor two-phase flow was proposed. The void fraction distribution in a fuel bundle under boiling heat transfer condition was analyzed and the bubble dynamics around the fuel rod surface were predicted quantitatively. (author)

  18. Thermal hydraulic issues and challenges for current and new generation FBRs

    International Nuclear Information System (INIS)

    In Liquid Metal Fast Breeder Reactors, especially Sodium Fast Reactors (SFR), thermal hydraulics poses several unique and challenging issues to be resolved for the design and safety analysis. For SFR, large heat transfer coefficient and high boiling point (888℃) are favorable to have a compact reactor working at low design pressures. However, the high heat transfer coefficient of sodium enhances the effects of thermal striping and also leads to large transient temperature variations in the adjoining structures following the thermal transients. This apart, particularly in pool type SFR where both hot and cold pools co-exist with large temperature differences, large volumetric expansion coefficient of sodium enhances the severity by introducing special phenomena such as thermal stratification, free level fluctuations and the associated temperature fluctuations in the structures. The demonstration of safe and reliable operation under all the operating conditions including prolonged station blackout calls for complex thermal hydraulics studies addressing forced/natural/mixed convection phenomena with multi-scale Computational Fluid Dynamics (CFD) codes. For the severe accident analysis, thermal hydraulics simulations are essential. For accomplishing these, thermal hydraulics studies of SFR demand state-of-art approaches and computer tools to get better insight and confidence on the reactor design. In the evolutionary design, coupled analysis for reactor physics, thermal hydraulics and structural mechanics analysis is carried out exploiting the computational capability fully (e.g. open source codes). Difficulties encountered in performing large-scale sodium experiments and inability of water to simulate thermal aspects of sodium, motivate the development of advanced numerical simulation techniques. Further, the functionality of some of the passive safety features is demonstrated with extensive numerical simulations; otherwise it is practically impossible to

  19. Whole core pin-by-pin coupled neutronic-thermal-hydraulic steady state and transient calculations using COBAYA3 code

    International Nuclear Information System (INIS)

    Nowadays coupled 3-dimensional neutron kinetics and thermal-hydraulic core calculations are performed by applying a radial average channel approach using a meshing of one quarter of assembly in the best case. This approach does not take into account the subchannels effects due to the averaging of the physical fields and the loose of heterogeneity in the thermal-hydraulic model. Therefore the models do not have enough resolution to predict those subchannels effects which are important for the fuel design safety margins, because it is in the local scale, where we can search the hottest pellet or the maximum heat flux. The Polytechnic University of Madrid advanced multi-scale neutron-kinetics and thermal-hydraulics methodologies being implemented in COBAYA3 include domain decomposition by alternate core dissections for the local 3-dimensional fine-mesh scale problems (pin cells/subchannels) and an analytical nodal diffusion solver for the coarse mesh scale coupled with the thermal-hydraulic using a model of one channel per assembly or per quarter of assembly. In this work, we address the domain decomposition by the alternate core dissections methodology applied to solve coupled 3-dimensional neutronic-thermal-hydraulic problems at the fine-mesh scale. The neutronic-thermal-hydraulic coupling at the cell-subchannel scale allows the treatment of the effects of the detailed thermal-hydraulic feedbacks on cross-sections, thus resulting in better estimates of the local safety margins at the pin level. (Author)

  20. European activities on crosscutting thermal-hydraulic phenomena for innovative nuclear systems

    International Nuclear Information System (INIS)

    Highlights: • This paper serves as a guidance of the special issue. • The technical tasks and methodologies applied to achieve the objectives have been described. • Main results achieved so far are summarized. - Abstract: Thermal-hydraulics is recognized as a key scientific subject in the development of innovative reactor systems. In Europe, a consortium is established consisting of 24 institutions of universities, research centers and nuclear industries with the main objectives to identify and to perform research activities on important crosscutting thermal-hydraulic issues encountered in various innovative nuclear systems. For this purpose the large-scale integrated research project THINS (Thermal-Hydraulics of Innovative Nuclear Systems) is launched in the 7th Framework Programme FP7 of European Union. The main topics considered in the THINS project are (a) advanced reactor core thermal-hydraulics, (b) single phase mixed convection, (c) single phase turbulence, (d) multiphase flow, and (e) numerical code coupling and qualification. The main objectives of the project are: • Generation of a data base for the development and validation of new models and codes describing the selected crosscutting thermal-hydraulic phenomena. • Development of new physical models and modeling approaches for more accurate description of the crosscutting thermal-hydraulic phenomena. • Improvement of the numerical engineering tools for the design analysis of the innovative nuclear systems. This paper describes the technical tasks and methodologies applied to achieve the objectives. Main results achieved so far are summarized. This paper serves also as a guidance of this special issue

  1. European activities on crosscutting thermal-hydraulic phenomena for innovative nuclear systems

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, X., E-mail: xu.cheng@kit.edu [Karlsruhe Institute of Technology (KIT) (Germany); Batta, A. [Karlsruhe Institute of Technology (KIT) (Germany); Bandini, G. [Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) (Italy); Roelofs, F. [Nuclear Research and Consultancy Group (NRG) (Netherlands); Van Tichelen, K. [Studiecentrum voor Kernenergie – Centre d’étude de l’Energie Nucléaire (SCK-CEN) (Belgium); Gerschenfeld, A. [Commissariat à l’Energie Atomique (CEA) (France); Prasser, M. [Paul Scherrer Institute (PSI) (Switzerland); Papukchiev, A. [Gesellschaft für Anlagen- und Reaktorsicherheit mbH (GRS) (Germany); Hampel, U. [Helmholtz-Zentrum Dresden-Rossendorf e.V. (HZDR) (Germany); Ma, W.M. [Kungliga Tekniska Högskolan (KTH) (Sweden)

    2015-08-15

    Highlights: • This paper serves as a guidance of the special issue. • The technical tasks and methodologies applied to achieve the objectives have been described. • Main results achieved so far are summarized. - Abstract: Thermal-hydraulics is recognized as a key scientific subject in the development of innovative reactor systems. In Europe, a consortium is established consisting of 24 institutions of universities, research centers and nuclear industries with the main objectives to identify and to perform research activities on important crosscutting thermal-hydraulic issues encountered in various innovative nuclear systems. For this purpose the large-scale integrated research project THINS (Thermal-Hydraulics of Innovative Nuclear Systems) is launched in the 7th Framework Programme FP7 of European Union. The main topics considered in the THINS project are (a) advanced reactor core thermal-hydraulics, (b) single phase mixed convection, (c) single phase turbulence, (d) multiphase flow, and (e) numerical code coupling and qualification. The main objectives of the project are: • Generation of a data base for the development and validation of new models and codes describing the selected crosscutting thermal-hydraulic phenomena. • Development of new physical models and modeling approaches for more accurate description of the crosscutting thermal-hydraulic phenomena. • Improvement of the numerical engineering tools for the design analysis of the innovative nuclear systems. This paper describes the technical tasks and methodologies applied to achieve the objectives. Main results achieved so far are summarized. This paper serves also as a guidance of this special issue.

  2. Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T [ORNL

    2014-05-01

    Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts

  3. Development of regulatory technology for thermal-hydraulic safety analysis

    International Nuclear Information System (INIS)

    The present study aims to develop the regulation capability in thermal-hydraulic safety analysis which was required for the reasonable safety regulation in the current NPP, the next generation reactors, and the future-type reactors. The fourth fiscal year of the first phase of the research was focused on the following research topics: Investigation on the current status of the thermal-hydraulic safety analysis technology outside and inside of the country; Review on the improved features of the thermal-hydraulic safety analysis regulatory audit code, RELAP5/MOD3; Assessments of code with LOFT L9-3 ATWS experiment and LSTF SB-SG-10 multiple SGTR experiment; Application of the RELAP5/CANDU code to analyses of SLB and LBLOCA and evaluation of its effect on safety; Application of the code to IAEA PHWR ISP analysis; Assessments of RELAP5 and TRAC with UPTF downcomer injection test and Analysis of LBLOCA with RELAP5 for the performance evaluation of KNGR DVI; Setup of a coupled 3-D kinetics and thermal-hydraulics and application it to a reactivity accident analysis; and Extension of database and improvement of plant input decks. For supporting the resolution of safety issues, loss of RHR event during midloop operation was analyzed for Kori Unit 3, issues on high burnup fuel were reviewed and performance of FRAPCON-3 assessed. Also MSLB was analyzed to figure out the sensitivity of downcomer temperature supporting the PTS risk evaluation of Kori Unit 1. Thermal stratification in pipe was analyzed using the method proposed. And a method predicting the thermal-hydraulic performance of IRWST of KNGR was explored. The PWR ECCS performance criteria was issued as a MOST Article 200-19.and a regulatory guide on evaluation methodology was improved to cover concerns raised from the related licensing review process

  4. Portable Life Support Subsystem Thermal Hydraulic Performance Analysis

    Science.gov (United States)

    Barnes, Bruce; Pinckney, John; Conger, Bruce

    2010-01-01

    This paper presents the current state of the thermal hydraulic modeling efforts being conducted for the Constellation Space Suit Element (CSSE) Portable Life Support Subsystem (PLSS). The goal of these efforts is to provide realistic simulations of the PLSS under various modes of operation. The PLSS thermal hydraulic model simulates the thermal, pressure, flow characteristics, and human thermal comfort related to the PLSS performance. This paper presents modeling approaches and assumptions as well as component model descriptions. Results from the models are presented that show PLSS operations at steady-state and transient conditions. Finally, conclusions and recommendations are offered that summarize results, identify PLSS design weaknesses uncovered during review of the analysis results, and propose areas for improvement to increase model fidelity and accuracy.

  5. Thermal and hydraulic analyses of the System 81 cold traps

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K.

    1977-06-15

    Thermal and hydraulic analyses of the System 81 Type I and II cold traps were completed except for thermal transients analysis. Results are evaluated, discussed, and reported. Analytical models were developed to determine the physical dimensions of the cold traps and to predict the performance. The FFTF cold trap crystallizer performances were simulated using the thermal model. This simulation shows that the analytical model developed predicts reasonably conservative temperatures. Pressure drop and sodium residence time calculations indicate that the present design will meet the requirements specified in the E-Specification. Steady state temperature data for the critical regions were generated to assess the magnitude of the thermal stress.

  6. Thermal-hydraulic methods in fast reactor safety

    International Nuclear Information System (INIS)

    Methods for the solution of thermal-hydraulic problems in liquid metal fast breeder reactors (LMFBRs) arising primarily from transient accident analysis are reviewed. Principal emphasis is given to the important phenomenological issues of sodium boiling and fuel motion. Descriptions of representative phenomenological and mathematical models, computational algorithms, advantages and limitations of the approaches, and current research needs and directions are provided

  7. Thermal-hydraulic research plan for Babcock and Wilcox plants

    International Nuclear Information System (INIS)

    This document presents a plan for thermal-hydraulic research for Babcock and Wilcox designed reactor systems. It describes the technical issues, regulatory needs, and the research necessary to address these needs. The plan also discusses the relationship between current and proposed research, and provides a tentative schedule to complete the required work

  8. Thermal-hydraulic design of the 200 MW NHR

    International Nuclear Information System (INIS)

    The main problems regarding the AST-500 NHR thermal-hydraulics are considered. Basic thermal data of the reactor plant are given and peculiarities of coolant parameters at natural convection in the primary circuit are discussed. The in-reactor instrumentation system is briefly describes, as well as the results of natural-convective flow characteristics investigations using reactor test models. (author). 4 refs, 5 figs

  9. On-Line Core Thermal-Hydraulic Model Improvement

    International Nuclear Information System (INIS)

    The objective of this project is to implement a fast-running 4-channel based code CETOP-D in an advanced reactor core protection calculator system(RCOPS). The part required for the on-line calculation of DNBR were extracted from the source of the CETOP-D code based on analysis of the CETOP-D code. The CETOP-D code was revised to maintain the input and output variables which are the same as in CPC DNBR module. Since the DNBR module performs a complex calculation, it is divided into sub-modules per major calculation step. The functional design requirements for the DNBR module is documented and the values of the database(DB) constants were decided. This project also developed a Fortran module(BEST) of the RCOPS Fortran Simulator and a computer code RCOPS-SDNBR to independently calculate DNBR. A test was also conducted to verify the functional design and DB of thermal-hydraulic model which is necessary to calculate the DNBR on-line in RCOPS. The DNBR margin is expected to increase by 2%-3% once the CETOP-D code is used to calculate the RCOPS DNBR. It should be noted that the final DNBR margin improvement could be determined in the future based on overall uncertainty analysis of the RCOPS

  10. Local thermal-hydraulic behaviour in tight 7-rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, X. [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Dongchuan Road 800, 200240 Shanghai (China); Institute for Nuclear and Energy Technologies, Research Centre Karlsruhe, Postfach 3640, 76021 Karlsruhe (Germany)], E-mail: chengxu@sjtu.edu.cn; Yu, Y.Q. [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Dongchuan Road 800, 200240 Shanghai (China)

    2009-10-15

    Advanced water-cooled reactor concepts with tight lattices have been proposed worldwide to improve the fuel utilization and the economic competitiveness. In the present work, experimental investigations were performed on thermal-hydraulic behaviour in tight hexagonal 7-rod bundles under both single-phase and two-phase conditions. Freon-12 was used as working fluid due to its convenient operating parameters. Tests were carried out under both single-phase and two-phase flow conditions. Rod surface temperatures are measured at a fixed axial elevation and in various circumferential positions. Test data with different radial power distributions are analyzed. Measured surface temperatures of unheated rods are used for the assessment of and comparison with numerical codes. In addition, numerical simulation using sub-channel analysis code MATRA and the computational fluid dynamics (CFD) code ANSYS-10 is carried out to understand the experimental data and to assess the validity of these codes in the prediction of flow and heat transfer behaviour in tight rod bundle geometries. Numerical results are compared with experimental data. A good agreement between the measured temperatures on the unheated rod surface and the CFD calculation is obtained. Both sub-channel analysis and CFD calculation indicates that the turbulent mixing in the tight rod bundle is significantly stronger than that computed with a well established correlation.

  11. Nuclear power plant thermal-hydraulic performance research program plan

    International Nuclear Information System (INIS)

    The purpose of this program plan is to present a more detailed description of the thermal-hydraulic research program than that provided in the NRC Five-Year Plan so that the research plan and objectives can be better understood and evaluated by the offices concerned. The plan is prepared by the Office of Nuclear Regulatory Research (RES) with input from the Office of Nuclear Reactor Regulation (NRR) and updated periodically. The plan covers the research sponsored by the Reactor and Plant Systems Branch and defines the major issues (related to thermal-hydraulic behavior in nuclear power plants) the NRC is seeking to resolve and provides plans for their resolution; relates the proposed research to these issues; defines the products needed to resolve these issues; provides a context that shows both the historical perspective and the relationship of individual projects to the overall objectives; and defines major interfaces with other disciplines (e.g., structural, risk, human factors, accident management, severe accident) needed for total resolution of some issues. This plan addresses the types of thermal-hydraulic transients that are normally considered in the regulatory process of licensing the current generation of light water reactors. This process is influenced by the regulatory requirements imposed by NRC and the consequent need for technical information that is supplied by RES through its contractors. Thus, most contractor programmatic work is administered by RES. Regulatory requirements involve the normal review of industry analyses of design basis accidents, as well as the understanding of abnormal occurrences in operating reactors. Since such transients often involve complex thermal-hydraulic interactions, a well-planned thermal-hydraulic research plan is needed

  12. Computer code application programme of TAEA for thermal hydraulic research

    International Nuclear Information System (INIS)

    Evaluation of thermal-hydraulic conditions, fuel behavior, and reactor kinetic during various operating and postulated accident conditions results in conclusions that support decision-making process, the review of license application, and the resolution of other technical issues related to nuclear safety. Also these activities increase the understanding and involvement into new technical developments. Thermal-hydraulic research activities at TAEA focus on the application of computer codes that simulate the behavior of the reactor system. The computer codes are used to analyzed loss of coolant accidents, and system transients in light water nuclear reactors and to assess the consequences if imbalance occurs and to determine the effectiveness of mitigating actions. TAEA has used nuclear reactor system codes (RELAP5/Mod3.2 and higher versions, PARCS) and nuclear plant visual analyzer codes (NPA, SNAP, and XNGR5) obtained by in the framework of the CAMP Agreement signed between TAEA and the United States Nuclear Regulatory Commission (US NRC). TAEA performs and documents the code assessments including improvements and error corrections. Moreover, research activities concerning the passive cooling application and simulations of advanced nuclear power plant have been carried out by both experimental and theoretical means. For example, the experiment test facility, which was designed to investigate the effect of noncondensable gases on condensation, was conducted in cooperation with the Mechanical Engineering Department of the Middle East Technical University in Ankara (Turkey) and was finished. The text matrix obtained from this research was also submitted to US NRC data bank. Application of RELAP5 code f system transients include International Standard Problem (ISP) studies (ISP 33, 38, 42, 45, 46), accident analysis for different reactors and special topics in nuclear heat transfer problems (mid-loop operation). Research studies of severe accidents assess the detailed

  13. Hydraulic characterization of aquifers by thermal response testing

    Science.gov (United States)

    Wagner, Valentin; Blum, Philipp; Bayer, Peter

    2014-05-01

    Temperature as a major physical quantity of the subsurface, and naturally occurring thermal anomalies are recognized as promising passive tracers to characterize the subsurface. Accelerated by the increasing popularity of geothermal energy, also active thermal field experiments have gained interest in hydrogeology. Such experiments involve artificial local ground heating or cooling. Among these, the thermal response test (TRT) is one of the most established field investigation techniques in shallow geothermal applications. It is a common method to investigate important subsurface heat transport parameters to design sustainable ground-source heat pump (GSHP) systems. During the test, the borehole heat exchanger (BHE) is heated up with a defined amount of energy by circulating a heat carrier fluid. By comparing temperature change between BHE inlet and outlet, the ability of the BHE to transfer heat or cold to the ambient ground is assessed. However, standard interpretation does not provide any insight into the governing processes of in-situ heat transfer. We utilize a groundwater advection sensitive TRT evaluation approach based on the analytical moving line source equation. It is shown that the TRT as a classical geothermal field test can also be used as a hydrogeological field test. Our approach benefits from the fact that thermal properties, such as thermal conductivity, of natural aquifers typically are much less variable than hydraulic properties, such as hydraulic conductivity. It is possible to determine a relatively small hydraulic conductivity range with our TRT evaluation approach, given realistic ranges for thermal conductivity, volumetric heat capacity, thermal dispersivity and thermal borehole resistance. The method is successfully tested on a large-scale geothermal laboratory experiment (9 m × 6 m × 4.5 m) and with a commercially performed TRT in the field scale. The laboratory experiment consists of a layered artificial aquifer, which is penetrated

  14. Study of thermal - hydraulic sensors signal fluctuations in PWR

    International Nuclear Information System (INIS)

    This thesis deals with signal fluctuations of thermal-hydraulic sensors in the main coolant primary of a pressurized water reactor. The aim of this work is to give a first response about the potentiality of use of these noise signals for the functionning monitoring. Two aspects have been studied: - the modelisation of temperature fluctuations of core thermocouples, by a Monte-Carlo method, gives the main characteristics of these signals and their domain of application. - the determination of eigenfrequency in the primary by an acoustic representation could permit the monitoring of local and global thermo-hydraulic conditions

  15. Space nuclear reactor SP-100 thermal-hydraulic simulation

    International Nuclear Information System (INIS)

    Since 1983 it has been under development in the USA the project SP-100 of space nuclear reactors for electric generation in a range of 100 to 1000 KWe. In this project the heat is generated at the core of a fast compact liquid lithium refrigerated reactor. Thermoelectric converters produce direct current electric energy and the primary and secondary loops flow is controlled by electromagnetic thermoelectric pumps (EMTE). In this work it is studied a system with a fast nuclear reactor, with similar characteristics to the SP-100, aiming at generating high electric power in space for a future application on the TERRA (Advanced Fast Reactor Technology) Project of IEAv (Institute for Advanced Studies). It will be presented the working principles, basic structure and operation characteristics of an electromagnetic thermoelectric pump (EMTE) for a liquid metal cooled nuclear reactor refrigeration loops flow control. In order to determine the operating point of the reactor, it is indispensable the simulation of the EMTE pump along with the other components of the system, once all the working parameters are connected. So, it has been developed a computer system, named BEMTE-3 (a FORTRAN micro-computer code), which simulates the primary and secondary refrigeration components of liquid metal cooled fast space reactor. This computer code also simulates the thermoelectric conversion, with the flow being controlled by the EMTE pump with thermoelectric converters, determining the system operation point for a given nominal operating power. The BEMTE-3 is used for the study of the SP-100 primary and secondary loops thermal-hydraulic simulation and for the calculation of the operating point of the system based on data from available projects. (author)

  16. Coupled Monte Carlo neutronics and thermal hydraulics for power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bernnat, W.; Buck, M.; Mattes, M. [Institut fuer Kernenergetik und Energiesysteme IKE, Universitaet Stuttgart, Pfaffenwaldring 31, D-70569 Stuttgart (Germany); Zwermann, W.; Pasichnyk, I.; Velkov, K. [Gesellschaft fuer Anlagen- und Reaktorsicherheit GRS MbH, Forschungszentrum, Boltzmannstrase 14, 85748 Garching (Germany)

    2012-07-01

    The availability of high performance computing resources enables more and more the use of detailed Monte Carlo models even for full core power reactors. The detailed structure of the core can be described by lattices, modeled by so-called repeated structures e.g. in Monte Carlo codes such as MCNP5 or MCNPX. For cores with mainly uniform material compositions, fuel and moderator temperatures, there is no problem in constructing core models. However, when the material composition and the temperatures vary strongly a huge number of different material cells must be described which complicate the input and in many cases exceed code or memory limits. The second problem arises with the preparation of corresponding temperature dependent cross sections and thermal scattering laws. Only if these problems can be solved, a realistic coupling of Monte Carlo neutronics with an appropriate thermal-hydraulics model is possible. In this paper a method for the treatment of detailed material and temperature distributions in MCNP5 is described based on user-specified internal functions which assign distinct elements of the core cells to material specifications (e.g. water density) and temperatures from a thermal-hydraulics code. The core grid itself can be described with a uniform material specification. The temperature dependency of cross sections and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. Applications will be shown for the stationary part of the Purdue PWR benchmark using ATHLET for thermal- hydraulics and for a generic Modular High Temperature reactor using THERMIX for thermal- hydraulics. (authors)

  17. FONESYS: The FOrum and NEtwork of SYStem Thermal-Hydraulic Codes in Nuclear Reactor Thermal-Hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, S.H., E-mail: k175ash@kins.re.kr [Korea Institute of Nuclear Safety (KINS) (Korea, Republic of); Aksan, N., E-mail: nusr.aksan@gmail.com [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Austregesilo, H., E-mail: henrique.austregesilo@grs.de [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) (Germany); Bestion, D., E-mail: dominique.bestion@cea.fr [Commissariat à l’énergie atomique et aux énergies alternatives (CEA) (France); Chung, B.D., E-mail: bdchung@kaeri.re.kr [Korea Atomic Energy Research Institute (KAERI) (Korea, Republic of); D’Auria, F., E-mail: f.dauria@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Emonot, P., E-mail: philippe.emonot@cea.fr [Commissariat à l’énergie atomique et aux énergies alternatives (CEA) (France); Gandrille, J.L., E-mail: jeanluc.gandrille@areva.com [AREVA NP (France); Hanninen, M., E-mail: markku.hanninen@vtt.fi [VTT Technical Research Centre of Finland (VTT) (Finland); Horvatović, I., E-mail: i.horvatovic@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Kim, K.D., E-mail: kdkim@kaeri.re.kr [Korea Atomic Energy Research Institute (KAERI) (Korea, Republic of); Kovtonyuk, A., E-mail: a.kovtonyuk@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Petruzzi, A., E-mail: a.petruzzi@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy)

    2015-01-15

    Highlights: • We briefly presented the project called Forum and Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics (FONESYS). • We presented FONESYS participants and their codes. • We explained FONESYS projects motivation, its main targets and working modalities. • We presented FONESYS position about projects topics and subtopics. - Abstract: The purpose of this article is to present briefly the project called Forum and Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics (FONESYS), its participants, the motivation for the project, its main targets and working modalities. System Thermal-Hydraulics (SYS-TH) codes, also as part of the Best Estimate Plus Uncertainty (BEPU) approaches, are expected to achieve a more-and-more relevant role in nuclear reactor technology, safety and design. Namely, the number of code-users can easily be predicted to increase in the countries where nuclear technology is exploited. Thus, the idea of establishing a forum and a network among the code developers and with possible extension to code users has started to have major importance and value. In this framework the FONESYS initiative has been created. The main targets of FONESYS are: • To promote the use of SYS-TH Codes and the application of the BEPU approaches. • To establish acceptable and recognized procedures and thresholds for Verification and Validation (V and V). • To create a common ground for discussing envisaged improvements in various areas, including user-interface, and the connection with other numerical tools, including Computational Fluid Dynamics (CFD) Codes.

  18. FONESYS: The FOrum and NEtwork of SYStem Thermal-Hydraulic Codes in Nuclear Reactor Thermal-Hydraulics

    International Nuclear Information System (INIS)

    Highlights: • We briefly presented the project called Forum and Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics (FONESYS). • We presented FONESYS participants and their codes. • We explained FONESYS projects motivation, its main targets and working modalities. • We presented FONESYS position about projects topics and subtopics. - Abstract: The purpose of this article is to present briefly the project called Forum and Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics (FONESYS), its participants, the motivation for the project, its main targets and working modalities. System Thermal-Hydraulics (SYS-TH) codes, also as part of the Best Estimate Plus Uncertainty (BEPU) approaches, are expected to achieve a more-and-more relevant role in nuclear reactor technology, safety and design. Namely, the number of code-users can easily be predicted to increase in the countries where nuclear technology is exploited. Thus, the idea of establishing a forum and a network among the code developers and with possible extension to code users has started to have major importance and value. In this framework the FONESYS initiative has been created. The main targets of FONESYS are: • To promote the use of SYS-TH Codes and the application of the BEPU approaches. • To establish acceptable and recognized procedures and thresholds for Verification and Validation (V and V). • To create a common ground for discussing envisaged improvements in various areas, including user-interface, and the connection with other numerical tools, including Computational Fluid Dynamics (CFD) Codes

  19. Hierarchic modeling of heat exchanger thermal hydraulics

    International Nuclear Information System (INIS)

    Volume Averaging Technique (VAT) is employed in order to model the heat exchanger cross-flow as a porous media flow. As the averaging of the transport equations lead to a closure problem, separate relations are introduced to model interphase momentum and heat transfer between fluid flow and the solid structure. The hierarchic modeling is used to calculate the local drag coefficient Cd as a function of Reynolds number Reh. For that purpose a separate model of REV is built and DNS of flow through REV is performed. The local values of heat transfer coefficient h are obtained from available literature. The geometry of the simulation domain and boundary conditions follow the geometry of the experimental test section used at U.C.L.A. The calculated temperature fields reveal that the geometry with denser pin-fins arrangement (HX1) heats fluid flow faster. The temperature field in the HX2 exhibits the formation of thermal boundary layer between pin-fins, which has a significant role in overall thermal performance of the heat exchanger. Although presented discrepancies of the whole-section drag coefficient Cd are large, we believe that hierarchic modeling is an appropriate strategy for calculation of complex transport phenomena in heat exchanger geometries.(author)

  20. Final design of a free-piston hydraulic advanced Stirling conversion system

    Science.gov (United States)

    Wallace, D. A.; Noble, J. E.; Emigh, S. G.; Ross, B. A.; Lehmann, G. A.

    1991-01-01

    Under the US Department of Energy's (DOEs) Solar Thermal Technology Program, Sandia National Laboratories is evaluating heat engines for solar distributed receiver systems. The final design is described of an engineering prototype advanced Stirling conversion system (ASCS) with a free-piston hydraulic engine output capable of delivering about 25 kW of electric power to a utility grid. The free-piston Stirling engine has the potential for a highly reliable engine with long life because it has only a few moving parts, has noncontacting bearings, and can be hermetically sealed. The ASCS is designed to deliver maximum power per year over a range of solar input with a design life of 30 years (60,000 h). The system includes a liquid Nak pool boiler heat transport system and a free-piston Stirling engine with high-pressure hydraulic output, coupled with a bent axis variable displacement hydraulic motor and a rotary induction generator.

  1. Final design of a free-piston hydraulic advanced Stirling conversion system

    Science.gov (United States)

    Wallace, D. A.; Noble, J. E.; Emigh, S. G.; Ross, B. A.; Lehmann, G. A.

    Under the US Department of Energy's (DOEs) Solar Thermal Technology Program, Sandia National Laboratories is evaluating heat engines for solar distributed receiver systems. The final design is described of an engineering prototype advanced Stirling conversion system (ASCS) with a free-piston hydraulic engine output capable of delivering about 25 kW of electric power to a utility grid. The free-piston Stirling engine has the potential for a highly reliable engine with long life because it has only a few moving parts, has noncontacting bearings, and can be hermetically sealed. The ASCS is designed to deliver maximum power per year over a range of solar input with a design life of 30 years (60,000 h). The system includes a liquid Nak pool boiler heat transport system and a free-piston Stirling engine with high-pressure hydraulic output, coupled with a bent axis variable displacement hydraulic motor and a rotary induction generator.

  2. Preliminary thermal hydraulic analysis of hyper fuel assembly using Matra

    International Nuclear Information System (INIS)

    Sub-channel analysis of HYPER fuel assembly was performed using MATRA which is a subchannel analysis code developed by KAERI based on COBRA-IV-I. The MATRA code was considered for comparison between codes and assessing the capability of overcoming the limitation of the SLTHEN code used in the previous works. Two types of single fuel assembly, i.e., average assembly and hot assembly were considered for the present work. The predicted peak cladding temperatures of the average and hot assemblies were 536,2 C and 653,8 C, respectively with the reference design parameters. The comparison of results obtained by two codes shows that there is a good agreement for the predicted thermal hydraulic behaviour. It is judged that MATRA as well as SLTHEN is a very useful tool for thermal hydraulic design of the HYPER core and MATRA can be used to make up for the limitation of SLTHEN. (author)

  3. Thermal-hydraulic analysis on reactor upper plenum of MONJU

    International Nuclear Information System (INIS)

    Thermal-hydraulics analyses of the reactor upper plenum of Monju, Japanese prototype of FBR, were performed in IAEA/Monju-CRP from 2008 to 2012. However, detail temperature and flow rate conditions of the inlets were required for an accurate analysis. In this paper we re-evaluated the inlet boundary condition (subassembly outlets) and performed another thermal-hydraulics analysis with Star-CCM+. The surface of the structures in the upper plenum was precisely modeled. The structures included a fuel transfer machine, in-vessel racks, flow-guide tubes, etc. The result was following: the structure didn't have large influence to the temperature distribution, and the analysis result of the temperature distribution on the thermocouple plug had some difference from the test result. (author)

  4. Thermal-hydraulic tests for reactor safety system

    International Nuclear Information System (INIS)

    Tests for the safety depressurization system, Sparger adopted for the Korean next generation reactor, APR1400 are carried out for several geometries with the B and C (Blowdown and Condensation) facility in the condition of high temperature and pressure and with a small test facility in the condition of atmospheric temperature and pressure. Tests for the critical heat flux are performed with the RCS(Reactor Coolant System) facility as well as with the Freon CHF Loop in the condition of high temperature and pressure. The atmospheric temperature and pressure facility is utilized for development of the high standard thermal hydraulic measurement technology. The optical method is developed to measure the local thermal-hydraulic behavior for the single and two-phase boiling phenomena

  5. Outage Risk Assessment and Management (ORAM) thermal-hydraulics toolkit

    International Nuclear Information System (INIS)

    A PC-based thermal-hydraulic toolkit for use in support of outage optimization, management and risk assessment has been developed. This mechanistic toolkit incorporates simple models of key thermal-hydraulic processes which occur during an outage, such as recovery from or mitigation of outage upsets; this includes heat-up of water pools following loss of shutdown cooling, inadvertent drain down of the RCS, boiloff of coolant inventory, heatup of the uncovered core, and reflux cooling. This paper provides a list of key toolkit elements, briefly describes the technical basis and presents illustrative results for RCS transient behavior during reflux cooling, peak clad temperatures for an uncovered core and RCS response to loss of shutdown cooling. (author)

  6. Thermal hydraulic model descrition of TASS/SMR

    International Nuclear Information System (INIS)

    The TASS/SMR code has been developed for the safety analysis of SMART. The governing equations were applied only to the primary coolant system in TASS which had been developed at KAERI. In TASS/SMR, the solution method is improved so that the primary and secondary coolant systems are solved simultaneously. Besides the solution method, thermal-hydraulic models are incorporated, in TASS/SMR, such as non-condensible gas model, helical steam generator heat transfer model, and passive residual heat removal system (PRHRS) heat transfer model for the application to SMART. The governing equtions of TASS/SMR are based on the drift-flux model so that the accidents and transients accompaning with two-phase flow can be analized. This report describes the governing equations and solution methods used in TASS/SMR and also includes the description for the thermal hydraulic models for SMART design

  7. Thermal-hydraulic modeling of reactivity accidents in MTR reactors

    Directory of Open Access Journals (Sweden)

    Khater Hany

    2006-01-01

    Full Text Available This paper describes the development of a dynamic model for the thermal-hydraulic analysis of MTR research reactors during a reactivity insertion accident. The model is formulated for coupling reactor kinetics with feedback reactivity and reactor core thermal-hydraulics. To represent the reactor core, two types of channels are considered, average and hot channels. The developed computer program is compiled and executed on a personal computer, using the FORTRAN language. The model is validated by safety-related benchmark calculations for MTR-TYPE reactors of IAEA 10 MW generic reactor for both slow and fast reactivity insertion transients. A good agreement is shown between the present model and the benchmark calculations. Then, the model is used for simulating the uncontrolled withdrawal of a control rod of an ETRR-2 reactor in transient with over power scram trip. The model results for ETRR-2 are analyzed and discussed.

  8. Thermal hydraulic model descrition of TASS/SMR

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Han Young; Kim, H. C.; Chung, Y. J.; Lim, H. S.; Yang, S. H

    2001-04-01

    The TASS/SMR code has been developed for the safety analysis of SMART. The governing equations were applied only to the primary coolant system in TASS which had been developed at KAERI. In TASS/SMR, the solution method is improved so that the primary and secondary coolant systems are solved simultaneously. Besides the solution method, thermal-hydraulic models are incorporated, in TASS/SMR, such as non-condensible gas model, helical steam generator heat transfer model, and passive residual heat removal system (PRHRS) heat transfer model for the application to SMART. The governing equtions of TASS/SMR are based on the drift-flux model so that the accidents and transients accompaning with two-phase flow can be analized. This report describes the governing equations and solution methods used in TASS/SMR and also includes the description for the thermal hydraulic models for SMART design.

  9. Thermal hydraulic analysis of Alfred bayonet tube steam generator

    OpenAIRE

    Caramello, Marco; Panella, Bruno; De Salve, Mario; Bertani, Cristina

    2015-01-01

    The paper analyzes the performance of ALFRED steam generator from the thermal-hydraulic point of view highlighting the effect of some design features. The parameters object of the study are the regenerative heat transfer, the dimension of the inner tube and the length of the bayonet. The system code RELAP5-3D/2.4.2 has been chosen for the analysis. Sensitivities analysis allowed the determination of the different design parameters influence, here briefly summarized. The increase of regenerati...

  10. The analysis of thermal-hydraulic models in MELCOR code

    International Nuclear Information System (INIS)

    The objective of the present work is to verify the prediction and analysis capability of MELCOR code about the progression of severe accidents in light water reactor and also to evaluate appropriateness of thermal-hydraulic models used in MELCOR code. Comparing the results of experiment and calculation with MELCOR code is carried out to achieve the above objective. Specially, the comparison between the CORA-13 experiment and the MELCOR code calculation was performed

  11. Thermal-hydraulic interfacing code modules for CANDU reactors

    International Nuclear Information System (INIS)

    The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis

  12. Application of thermal-hydraulic codes in the nuclear sector

    International Nuclear Information System (INIS)

    Use of thermal-hydraulic codes is extended all over many different aspects of nuclear engineering. This article groups and briefly describes the main features of some of the well known codes as an introduction to their recent applications in the Spain nuclear sector. the broad range and quality of applications highlight the maturity achieved both in industry and research organizations and universities within the Spanish nuclear sector. (Author)

  13. Thermal-hydraulic interfacing code modules for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, W.S.; Gold, M.; Sills, H. [Ontario Hydro Nuclear, Toronto (Canada)] [and others

    1997-07-01

    The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

  14. Advanced Thermally Stable Jet Fuels

    Energy Technology Data Exchange (ETDEWEB)

    A. Boehman; C. Song; H. H. Schobert; M. M. Coleman; P. G. Hatcher; S. Eser

    1998-01-01

    The Penn State program in advanced thermally stable jet fuels has five components: 1) development of mechanisms of degradation and solids formation; 2) quantitative measurement of growth of sub-micrometer and micrometer-sized particles during thermal stressing; 3) characterization of carbonaceous deposits by various instrumental and microscopic methods; 4) elucidation of the role of additives in retarding the formation of carbonaceous solids; and 5) assessment of the potential of producing high yields of cycloalkanes and hydroaromatics from coal.

  15. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

    Energy Technology Data Exchange (ETDEWEB)

    Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

    1995-09-01

    This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  16. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

    International Nuclear Information System (INIS)

    This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  17. Development of demonstration advanced thermal reactor

    International Nuclear Information System (INIS)

    The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported. (Kako, I.)

  18. Development of realistic thermal-hydraulic system analysis codes ; development of thermal hydraulic test requirements for multidimensional flow modeling

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Yull; Yoon, Sang Hyuk; Noh, Sang Woo; Lee, Il Suk [Seoul National University, Seoul (Korea)

    2002-03-01

    This study is concerned with developing a multidimensional flow model required for the system analysis code MARS to more mechanistically simulate a variety of thermal hydraulic phenomena in the nuclear stem supply system. The capability of the MARS code as a thermal hydraulic analysis tool for optimized system design can be expanded by improving the current calculational methods and adding new models. In this study the relevant literature was surveyed on the multidimensional flow models that may potentially be applied to the multidimensional analysis code. Research items were critically reviewed and suggested to better predict the multidimensional thermal hydraulic behavior and to identify test requirements. A small-scale preliminary test was performed in the downcomer formed by two vertical plates to analyze multidimensional flow pattern in a simple geometry. The experimental result may be applied to the code for analysis of the fluid impingement to the reactor downcomer wall. Also, data were collected to find out the controlling parameters for the one-dimensional and multidimensional flow behavior. 22 refs., 40 figs., 7 tabs. (Author)

  19. Thermal hydraulic simulations, error estimation and parameter sensitivity studies in Drekar::CFD

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Thomas Michael [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Shadid, John N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Pawlowski, Roger P. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cyr, Eric C. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wildey, Timothy Michael [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-01-01

    This report describes work directed towards completion of the Thermal Hydraulics Methods (THM) CFD Level 3 Milestone THM.CFD.P7.05 for the Consortium for Advanced Simulation of Light Water Reactors (CASL) Nuclear Hub effort. The focus of this milestone was to demonstrate the thermal hydraulics and adjoint based error estimation and parameter sensitivity capabilities in the CFD code called Drekar::CFD. This milestone builds upon the capabilities demonstrated in three earlier milestones; THM.CFD.P4.02 [12], completed March, 31, 2012, THM.CFD.P5.01 [15] completed June 30, 2012 and THM.CFD.P5.01 [11] completed on October 31, 2012.

  20. Scaling philosophy and system description of AHWR Thermal-Hydraulic Test Facility (ATTF)

    International Nuclear Information System (INIS)

    The Advanced Heavy Water Reactor (AHWR) being designed in India is a 920 MWth pressure tube type boiling light water cooled and heavy water moderated reactor. AHWR Thermal Hydraulic Test Facility (ATTF), a scaled experimental facility that simulates the thermal-hydraulic behaviour of main heat transport system and ECCS, is designed. The objectives of the facility are to obtain thermal margin (CHF) and the parallel channel stability behaviour Global scaling is based on Power to Volume ratio. This philosophy is based on maintaining the same pressure, temperature with same working fluid. Main advantage of this scaling approach is that it preserves the time scales which are very crucial for the simulation of transient and accident conditions to assess the performance of safety systems. All of the Main Heat Transport (MHT) and Emergency Core Cooling System (ECCS) components are scaled down on the basis of power to volume scaling. ATTF contains two full power channels in comparison with 452 channels of AHWR then the scaling ratio is 226. Therefore the volumes of the components in natural circulation path (MHT) are scaled down by 226. Different local phenomenon like Critical Heat Flux (CHF), Flashing, Geysering etc which affects the performance of the system are scaled down appropriately. GDCS injection, feed water flow etc are simulated as boundary flow scaling approach. This 3-level approach simulates almost all the thermal hydraulics phenomenon of the prototype in the model, with the appropriate scale of the model to the prototype. (author)

  1. Operating experience of reactors points up need for new thermal-hydraulic inquiries

    International Nuclear Information System (INIS)

    Review of accident and preaccident situation in the context of thermal-hydraulic processes in PWR and BWR is presented. The most frequently occurring preaccident events in the reactor operation pertaining to thermal-hydraulic processes: water hammer, thermal fatigue, transition processes, supercooling, formation of vortex, oscillation of power in BWR are discussed. Activation of theoretical and experimental thermal-hydraulic studies with the aim of improvement of safety and efficiency of NPU is proposed

  2. A two-compartment thermal-hydraulic experiment (LACE-LA4) analyzed by ESCADRE code

    International Nuclear Information System (INIS)

    Large scale experiments show that whenever a Loss of Coolant Accident (LOCA) occurs, water pools are generated. Stratifications of steam saturated gas develop above water pools causing a two-compartment thermal-hydraulics. The LACE (LWR Advanced Containment Experiment) LA4 experiment, performed at the Hanford Engineering Development Laboratory (HEDL), exhibited a strong stratification, at all times, above a growing water pool. JERICHO and AEROSOLS-B2 are part of the ESCADRE code system (Ensemble de Systemes de Codes d'Analyse d'accident Des Reacteurs A Eau), a tool for evaluating the response of a nuclear plant to severe accidents. These two codes are here used to simulate respectively the thermal-hydraulics and the associated aerosol behavior. Code results have shown that modelling large containment thermal-hydraulics without taking account of the stratification phenomenon leads to large overpredictions of containment pressure and temperature. If the stratification is modelled as a zone with a higher steam condensation rate and a higher thermal resistance, ESCADRE predictions match quite well experimental data. The stratification thermal-hydraulics is controlled by power (heat fluxes) repartition in the lower compartment between the water pool and the nearby walls. Therefore the total, direct heat exchange between the two compartment is reduced. Stratification modelling is believed to be important for its influence on aerosol behavior: aerosol deposition through the inter-face of the two subcompartments is improved by diffusiophoresis and thermophoresis. In addition the aerosol concentration gradient, through the stratification, will cause a driving force for motion of smaller particles towards the pool. (author)

  3. Thermal Hydraulic Analysis on Containment Filtered Venting System

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Suk; Park, Tong Kyu; Lee, Doo Yong; Lee, Byung Chul [FNC Technology Co. Ltd., Yongin (Korea, Republic of); Lee, Sang Won; Kim, Hyeong Taek [KHNP-Central Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In this study, the thermal hydraulic conditions (e. g. pressure and flow rate) at each component have been examined and the sensitivity analysis on CFVS design parameters (e. g. water inventory, volumetric flow rate). The purpose is to know the possible range of flow conditions at each component to determine the optimum size of filtration system. GOTHIC code has been used to simulate the thermal-hydraulic behavior inside of CFVS. The behavior of flows in the CFVS has been investigated. The vessel water level and the flow rates during the CFVS operation are examined. It was observed that the vessel water level would be changed significantly due to steam condensation/thermal expansion and steam evaporation. Therefore, the vessel size and the initial water inventory should be carefully determined to keep the minimum water level required for filtration components and not to flood the components in the upper side of the vessel. It has been also observed that the volumetric flow rate is maintained during the CFVS operation, which is beneficial for pool scrubbing units. However, regarding the significant variations at the orifice downstream, careful design would be necessary.

  4. Simplified model of VVER–1000 thermal hydraulic process

    International Nuclear Information System (INIS)

    This report introduce developed mathematical model of thermal hydraulic process which occurs in the core of VVER – 1000 type of nuclear reactor and the coolant flow is considered in one dimension. Navier – Stokes differential equations system is taken like basis – namely continuity equation and momentum and energy conservation equations with two algebraic equations of state by which closure relationship is done. Following the approach of simplifying the model in momentum and energy equations some simplifying assumptions is made like Reynolds term in momentum equation is neglected and diffusion term – in energy equation. The differential equations system can be split into two parts: at one hand continuity equation and momentum equation are solved regarding velocity and pressure distribution and at another – solving energy equation. This report is considering the second case – solving the unsteady energy equation at prescribed distributions of velocity and pressure. By using one of the algebraic state equation, i=cvT, the energy equation is written regarding the temperature. Velocity and pressure given in the model are estimated by the thermal hydraulics means. The energy equation is solved by finite volumes method at which considered region is divided by N finite volumes, scalar values T and P are represented at central points at volumes and the velocity – at borders of these volumes by so called staggered grid. The equation is integrated at the boundaries of every finite volume and in time at the interval [t; t+ t]. For results obtained by integration is applied Crank-Nicolson semi-implicit scheme. As a result this gives an algebraic system with three diagonal matrix which can be solved with Crout effective algorithm. Keywords: finite volumes method, RELAP5, thermal hydraulics

  5. Thermal hydraulic design of the mercury target vessel

    International Nuclear Information System (INIS)

    Japan Atomic Energy Agency (JAEA) has completed the mercury target system as the pulsed spallation neutron source of J-PARC project, which has the highest level power of 1MW in the world. The basic design of the flow channel and structure of the mercury target vessel, which is the core of the neutron source, was carried out by JAEA, and the detail design, parts fabrication and assembling have been carried out by the vendor from 2003. Taking these fabrication designs and assembling conditions into consideration, the final performance evaluation of the mercury target vessel was carried out from the viewpoint of thermal hydraulics. The general thermal hydraulic analyses code, STAR-CD, was used, and the thermal hydraulic analyses were carried out for the mercury flow in the mercury vessel and the heavy water flow in the safety hull, taking the nuclear heating and the heat transportation into consideration. The analytical model was three dimensional. The total cell number of the mercury vessel and mercury was 1.78 x 106 and that of the safety hull was 2.39 x 106. The standard k-ε model and MARS were adopted as the basic combination of the turbulence model and the differential scheme, but other combinations, such as RNG k-ε model and UD were also used as a reference. Comparing these analytical results, it was confirmed that the mercury target vessel fulfills the design requirements such as the fluid inlet velocity, the maximum temperature of fluid, the maximum temperature of the vessel, the pressure drop of fluid, etc. The influence of the welding deformation of the mercury target vessel was also evaluated. Mercury flow and heavy water flow are affected a little, but they do not much extend the required condition, and the structural integrity was confirmed. (author)

  6. Advanced solar thermal receiver technology

    Science.gov (United States)

    Kudirka, A. A.; Leibowitz, L. P.

    1980-01-01

    Development of advanced receiver technology for solar thermal receivers designed for electric power generation or for industrial applications, such as fuels and chemical production or industrial process heat, is described. The development of this technology is focused on receivers that operate from 1000 F to 3000 F and above. Development strategy is mapped in terms of application requirements, and the related system and technical requirements. Receiver performance requirements and current development efforts are covered for five classes of receiver applications: high temperature, advanced Brayton, Stirling, and Rankine cycle engines, and fuels and chemicals.

  7. Current Development and Trends in Thermal-Hydraulics

    International Nuclear Information System (INIS)

    A review of CSNI activities during the last two decades in the field of thermal-hydraulics and related topics has been extensively presented in sessions 2. to 9. New activities are in progress or planned partly based on recommendations of the CSNI Operating Plan and the CSNI SESAR SFEAR report, but also on requests coming from the member states. These activities are performed in the frame of the CSNI Working Group on the Analysis and Management of Accidents (GAMA) or in the frame of CSNI Projects. These actions are summarized in this paper.

  8. 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)

  9. Fractured Apache leap tuff: Interstitial, hydraulic, pneumatic, and thermal properties

    International Nuclear Information System (INIS)

    Unsaturated fractured tuff characterization with respect to flow and transport properties presents substantial methodological and technological challenges to the scientific and engineering communities. Methods and techniques are now limited to saturated and shallow unsaturated soil media. Application of these methods and techniques to deep unsaturated fractured rock media requires that existing methods be verified, as well as the development and application of new and imaginative procedures and equipment when the existing methods are inadequate. This paper presents laboratory and field data which are used for interpreting characterization methods. Interstitial, hydraulic, pneumatic, and thermal data sets are presented over a wide range of water contents and matric suctions. 14 refs., 12 tabs

  10. Steady thermal hydraulic analysis for a molten salt reactor

    Institute of Scientific and Technical Information of China (English)

    ZHANG Dalin; QIU Suizheng; LIU Changliang; SU Guanghui

    2008-01-01

    The Molten Salt Reactor (MSR) can meet the demand of transmutation and breeding. In this study, theoretical calculation of steady thermal hydraulic characteristics of a graphite-moderated channel type MSR is conducted. The DRAGON code is adopted to calculate the axial and radial power factor firstly. The flow and heat transfer model in the fuel salt and graphite are developed on basis of the fundamental mass, momentum and energy equations. The results show the detailed flow distribution in the core, and the temperature profiles of the fuel salt, inner and outer wall in the nine typical elements along the axial flow direction are also obtained.

  11. Thermal hydraulic behavior of SCWR sliding pressure startup

    International Nuclear Information System (INIS)

    The modification to ATHLET-SC code is introduced in this paper, which realizes the simulation of trans-critical transients using two-phase model. With the modified code, the thermal-hydraulic dynamic behavior of the mixed SCWR core during the startup process is simulated. The startup process is similar to the design of SCLWR-H sliding pressure startup. The results show that maximum temperature of cladding-surface does not exceed 650℃ in the whole startup process, and the sudden change of water properties in the trans-critical transients will not cause harmful influence to the heat transfer of the fuel cladding. (authors)

  12. Thermal-hydraulic modeling of reracked spent fuel pool

    International Nuclear Information System (INIS)

    The simple model of the spent fuel pool for computer code GOTHIC, which enables calculation of thermal-hydraulic parameters of the reracked spent fuel pool of NPP Krsko, has been developed. This model encompasses all basic characteristics of spent fuel pool, which are necessary to simulate a global behavior of spent fuel pool cooling. Within this model, the temperatures of the spent fuel pool for steady state, as well as temperature increases after loss of cooling were calculated for NPP Krsko reracked spent fuel pool. (author)

  13. Basis for Coupled 3-D Neutronics-Thermal-Hydraulics

    OpenAIRE

    Aragonés Beltrán, José María

    2008-01-01

    The purpose of this seminar is first to discuss the basis of the coupling between 3-D Neutron- Kinetics and Thermal-Hydraulics codes, including the control and 3-D variables to interchange, the transform of the 3-D NK and TH core nodalizations, and the schemes for temporal coupling and time-step control. As representative examples of the NK-TH core coupling, we discuss first the integration of a 3-D NK nodal code with a TH subchannel code, for detailed transient core analysis; and second the ...

  14. Thermal-hydraulic analysis for reactor vessel upper-head small break LOCA using SPACE code

    International Nuclear Information System (INIS)

    A small break loss of coolant accident (SBLOCA) in upper-head of a reactor vessel at OPR1000 was analyzed using SPACE code, which is an advanced thermal-hydraulic system analysis code developed by the Korea nuclear industry. To assess the capability of SPACE code, upper-head SBLOCA with full plant safeguards was simulated, and compared with results of MARS-KS code. Reasonably good agreement with major thermal-hydraulic parameters was obtained by analyzing the transient behavior. Based on the observed thermal-hydraulic features, simulations with the failure of partial plant safeguards were conducted to analyze the safety and performance of OPR1000. Effects of failure to scram and high-pressure safety injection (HPSI) were investigated, and safety assessment was evaluated according to operator actions. Comparative study without any emergency core cooling systems (ECCS) was also conducted to judge the severity of the break location. From the results, this indicated that SPACE code has capabilities to simulate upper-head SBLOCA, and OPR1000 was evaluated to have sufficient safety margin with the application of proper emergency operating procedures.

  15. A Parametric Study on the Thermal Hydraulic Design for an Annular Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Shin, C. H.; Seo, K. W.; Chun, T. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-10-15

    Recently, MIT proposed an internally and externally cooled annular fuel for an advanced PWR which can endure a substantial power uprating. To apply this annular fuel in the conventional reactors such as OPR-1000, it is desirable to investigate its a structural compatibility for its reloading to operating PWR reactors of OPR-1000 as well as other compatibilities like the fuel to moderator ratio, amount of fissile material and coolant flow area. Conventional fuel assembly has a 16x16 solid rod array with four big guide tubes and one instrumentation tube. A 12x12 annular fuel assembly design which can meet the above compatibilities was proposed, which is structurally compatible with the existing internals of OPR-1000. Actually the advantage of an annular fuel comes from the fuel performance and thermal hydraulics. In the thermal hydraulic analysis, the mixing effect between the neighboring channels has to be carried out in a subchannel analysis. A subchannel analysis code, MATRA has been developed by KAERI. However, MATRA dose not have the capability to model both an internally and externally cooled annular fuel. A subchannel code, MATRA-AF which can be coupled to MATRA and can calculate the coolant flow distribution and heat transfer fraction in the internal and external subchannels has been developed. In this paper, the characteristics and the verification of the MATRA-AF are described. The effects of the thermal hydraulic parameters are estimated through a single fuel assembly.

  16. A Parametric Study on the Thermal Hydraulic Design for an Annular Fuel Assembly

    International Nuclear Information System (INIS)

    Recently, MIT proposed an internally and externally cooled annular fuel for an advanced PWR which can endure a substantial power uprating. To apply this annular fuel in the conventional reactors such as OPR-1000, it is desirable to investigate its a structural compatibility for its reloading to operating PWR reactors of OPR-1000 as well as other compatibilities like the fuel to moderator ratio, amount of fissile material and coolant flow area. Conventional fuel assembly has a 16x16 solid rod array with four big guide tubes and one instrumentation tube. A 12x12 annular fuel assembly design which can meet the above compatibilities was proposed, which is structurally compatible with the existing internals of OPR-1000. Actually the advantage of an annular fuel comes from the fuel performance and thermal hydraulics. In the thermal hydraulic analysis, the mixing effect between the neighboring channels has to be carried out in a subchannel analysis. A subchannel analysis code, MATRA has been developed by KAERI. However, MATRA dose not have the capability to model both an internally and externally cooled annular fuel. A subchannel code, MATRA-AF which can be coupled to MATRA and can calculate the coolant flow distribution and heat transfer fraction in the internal and external subchannels has been developed. In this paper, the characteristics and the verification of the MATRA-AF are described. The effects of the thermal hydraulic parameters are estimated through a single fuel assembly

  17. Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures

    Directory of Open Access Journals (Sweden)

    Alessandro Petruzzi

    2008-01-01

    Full Text Available In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first system codes, developed at the beginning of the 1970s, utilized the homogenous equilibrium model with three balance equations to describe the two-phase flow, nowadays the more advanced system codes are based on the so-called “two-fluid model” with separation of the water and vapor phases, resulting in systems with at least six balance equations. The wide experimental campaign, constituted by the integral and separate effect tests, conducted under the umbrella of the OECD/CSNI was at the basis of the development and validation of the thermal-hydraulic system codes by which they have reached the present high degree of maturity. However, notwithstanding the huge amounts of financial and human resources invested, the results predicted by the code are still affected by errors whose origins can be attributed to several reasons as model deficiencies, approximations in the numerical solution, nodalization effects, and imperfect knowledge of boundary and initial conditions. In this context, the existence of qualified procedures for a consistent application of qualified thermal-hydraulic system code is necessary and implies the drawing up of specific criteria through which the code-user, the nodalization, and finally the transient results are qualified.

  18. Benchmark of Neutronics and Thermal-hydraulics Coupled Simulation program NTC on beam interruptions in XADS

    International Nuclear Information System (INIS)

    Highlights: • A Neutronics and Thermal-hydraulics Coupled code is developed for transient analysis. • The spatial kinetics model was employed in the benchmark. • The simulation correctness of NTC accuracy demonstrated by benchmark. - Abstract: The Neutronics and Thermal-hydraulics Coupled Simulation program (NTC) is developed by FDS Team, which is a code used for transient analysis of advanced reactors. To investigate the capacity and calculation correctness of NTC for transient simulation, a benchmark on beam interruptions in an 80 MWth LBE-cooled and MOX-fuelled experimental accelerator-driven sub-critical system XADS was carried out by NTC. The benchmark on beam interruptions used in this paper was developed by the OECD/NEA Working Party on Scientific Issues in Partitioning and Transmutation (WPPT). The calculation model had the minimum phenomenological and computational complexity which concerned a simple model (single fuel channel thermal-hydraulics) of the average fuel pin corresponding to the BOL fuel condition. This benchmark was designed to investigate the temperature and power responses caused by beam interruption of different durations, which aimed at comparative assessment of NTC and other computation methods. A comparison of NTC and other ten sets of temperature and power was provided, which showed that the results had good agreement

  19. Thermal-Hydraulic System Codes in Nuclear Reactor Safety and Qualification Procedures

    International Nuclear Information System (INIS)

    In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first system codes, developed at the beginning of the 1970s, utilized the homogenous equilibrium model with three balance equations to describe the two-phase flow, nowadays the more advanced system codes are based on the so-called two-fluid model with separation of the water and vapor phases, resulting in systems with at least six balance equations. The wide experimental campaign, constituted by the integral and separate effect tests, conducted under the umbrella of the OECD/CSNI was at the basis of the development and validation of the thermal-hydraulic system codes by which they have reached the present high degree of maturity. However, notwithstanding the huge amounts of financial and human resources invested, the results predicted by the code are still affected by errors whose origins can be attributed to several reasons as model deficiencies, approximations in the numerical solution, nodalization effects, and imperfect knowledge of boundary and initial conditions. In this context, the existence of qualified procedures for a consistent application of qualified thermal-hydraulic system code is necessary and implies the drawing up of specific criteria through which the code-user, the nodalization, and finally the transient results are qualified

  20. Thermal-hydraulic analysis for reactor vessel upper-head small break LOCA using SPACE code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minhee; Kim, Seyun [Korea Hydro and Nuclear Power Co., Central Research Inst., Daejeon (Korea, Republic of)

    2015-08-15

    A small break loss of coolant accident (SBLOCA) in upper-head of a reactor vessel at OPR1000 was analyzed using SPACE code, which is an advanced thermal-hydraulic system analysis code developed by the Korea nuclear industry. To assess the capability of SPACE code, upper-head SBLOCA with full plant safeguards was simulated, and compared with results of MARS-KS code. Reasonably good agreement with major thermal-hydraulic parameters was obtained by analyzing the transient behavior. Based on the observed thermal-hydraulic features, simulations with the failure of partial plant safeguards were conducted to analyze the safety and performance of OPR1000. Effects of failure to scram and high-pressure safety injection (HPSI) were investigated, and safety assessment was evaluated according to operator actions. Comparative study without any emergency core cooling systems (ECCS) was also conducted to judge the severity of the break location. From the results, this indicated that SPACE code has capabilities to simulate upper-head SBLOCA, and OPR1000 was evaluated to have sufficient safety margin with the application of proper emergency operating procedures.

  1. Hydraulics.

    Science.gov (United States)

    Decker, Robert L.; Kirby, Klane

    This curriculum guide contains a course in hydraulics to train entry-level workers for automotive mechanics and other fields that utilize hydraulics. The module contains 14 instructional units that cover the following topics: (1) introduction to hydraulics; (2) fundamentals of hydraulics; (3) reservoirs; (4) lines, fittings, and couplers; (5)…

  2. Minerve: thermal-hydraulic phenomena simulation and virtual reality

    International Nuclear Information System (INIS)

    MINERVE is a 3D interactive application representing the thermal-hydraulic phenomena happening in a nuclear plant. Therefore, the 3D geometric model of the French 900 MW PWR installations has been built. The users can interact in real time with this model to see at each step of the simulation what happens in the pipes. The thermal-hydraulic simulation is made by CATHARE-2, which calculates at every time step data on about one thousand meshes (the whole primary circuit, a part of the second circuit, and the Residual Heat Removal System). The simulation covers incidental and accidental cases on these systems. There are two main innovations in MINERVE: In the domain of nuclear plant's visualization, it is to introduce interactive 3D software mechanisms to visualize results of a physical simulation. In the domain of real-time 3D, it is to visualize fluids in a pipe, while they can have several configurations, like bubbles or single liquid phase. These mechanisms enable better comprehension and better visual representation of the possible phenomena. This paper describes the functionalities of MINERVE, and the difficulties to represent fluids with several characteristics like speed, configuration,..., in 3D. On the end, we talk about the future of MINERVE, and more widely of the possible futures of such an application in scientific visualization. (authors)

  3. Thermal hydraulic and safety analysis for Tajoura Research Center

    International Nuclear Information System (INIS)

    Thermal hydraulic and safety analysis of Tajoura Research Center (TRR) utilizing low enriched uranium (LEU) fuel type IRT-4M have been performed using computer code PARET. The compact loading of the present core comprises of 16 fuel assemblies and 11 control rods. Results of the thermal hydraulic analysis show that the reactor can be operated at steady-state power of 10 MW for a flow rate of 533 m3/h, with sufficient margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions, the considered events are, positive reactivity insertion, flow reduction due to loss of primary coolant. For each of these transients, time history of reactor power, energy releases, clad surface and fuel centerline temperatures and maximum heat flux ratios were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these transients. It is therefore concluded that the reactor can be operated at steady-state power level of 10 MW without compromising safety. (author)

  4. Thermal hydraulic analysis of ETRR-2 using RELAP5 code

    International Nuclear Information System (INIS)

    The present work was developed within the frame of the IAEA Coordinated Research Project 1496 ''Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal-hydraulic computational methods and tools for operation and safety analysis of research reactors''. Three benchmark experiments were designed and performed to study the performance of the Egyptian Research Reactor ETRR-2. The experiments included steady state measurements as well as loss of flow transient (LOFT) and loss of power transient (SCRAM) conditions. The Code RELAP5/Mod3.4 was used to simulate the components of the ETRR-2 systems for the thermal hydraulic analysis of the reactor. The dimensions and elevations of the primary cooling components are based on real conditions (3-D configuration). RELAP5 results provided benchmark data which verified the experimental measurements taken from instrumentations installed for this experiment at several positions in the core. The predicted values for the coolant and clad surface temperature at different locations in the core showed a remarkable agreement with the experimental values, for both the steady state and transient conditions.

  5. Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    Thermal hydraulic and safety analysis of Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel have been performed using computer code PARET. The present core comprises of 29 standard and 5 control fuel elements. Results of the thermal hydraulic analysis show that the core can be operated at a steady-state power level of 10 MW for a flow rate of 950 m3/h, with sufficient safety margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions. The events studied include: start-up accident; accidental drop of a fuel element in the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is therefore concluded that the reactor can be safely operated at 10 MW without compromising safety. (author)

  6. LWR containment thermal hydraulic codes benchmark demona B3 exercise

    International Nuclear Information System (INIS)

    Recent discussion about the aerosol codes currently used for the analysis of containment retention capabilities have revealed a number of questions concerning the reliabilities and verifications of the thermal-hydraulic modules of these codes with respect to the validity of implemented physical models and the stability and effectiveness of numerical schemes. Since these codes are used for the calculation of the Source Term for the assessment of radiological consequences of severe accidents, they are an important part of reactor safety evaluation. For this reason the Commission of European Communities (CEC), following the recommendation mode by experts from Member Stades, is promoting research in this field with the aim also of establishing and increasing collaboration among Research Organisations of member countries. In view of the results of the studies, the CEC has decided to carry out a Benchmark exercise for severe accident containment thermal hydraulics codes. This exercise is based on experiment B3 in the DEMONA programme. The main objective of the benchmark exercise has been to assess the ability of the participating codes to predict atmosphere saturation levels and bulk condensation rates under conditions similar to those predicted to follow a severe accident in a PWR. This exercise follows logically on from the LA-4 exercise, which, is related to an experiment with a simpler internal geometry. We present here the results obtained so far and from them preliminary conclusions are drawn, concerning condensation temperature, pressure, flow rates, in the reactor containment

  7. Parallelization of detailed thermal-hydraulic analysis program SPIRAL

    International Nuclear Information System (INIS)

    The detailed thermal-hydraulic analysis computer program APIRAL is under development for the evaluation of local flow and temperature fields in wire-wrapped fuel pin bundles deformed by the influence of high burn-up, which are hard to reveal by experiment due to measurement difficulty. The coupling utilization of this program and a subchannel analysis program can offer a practical method to evaluate thermal-hydraulic behavior in a whole fuel assembly with high accuracy. This report describes the parallelization of SPIRAL for improving applicability to larger numerical simulations. The domain decomposition method using overlapped elements was adopted to the parallelization because SPIRAL is based on finite element method and it can minimize the number of communications between processor elements. As a parallelization programming library, Massage Passing Interface (MPI) was applied. Several numerical simulations were carried out to verify the parallelized version of SPIRAL and to evaluate parallelization efficiency. From These simulation results, the validity of this version was confirmed. Although no good parallelization efficiency was obtained in the case of small scale simulations due to overhead processes, approximately twelve times processing speed was achieved by using 16 processor elements in larger scale simulations. (author)

  8. Proceedings of the 8. Brazilian Meeting on Reactor Physics and Thermal Hydraulics

    International Nuclear Information System (INIS)

    Some papers about pressurized light water reactors, fast reactors, accident analysis, transients, research reactors, nuclear data collection, thermal hydraulics, reactor monitoring, neutronics are presented. (E.G.)

  9. Advanced thermally stable jet fuels

    Energy Technology Data Exchange (ETDEWEB)

    Schobert, H.H.

    1999-01-31

    The Pennsylvania State University program in advanced thermally stable coal-based jet fuels has five broad objectives: (1) Development of mechanisms of degradation and solids formation; (2) Quantitative measurement of growth of sub-micrometer and micrometer-sized particles suspended in fuels during thermal stressing; (3) Characterization of carbonaceous deposits by various instrumental and microscopic methods; (4) Elucidation of the role of additives in retarding the formation of carbonaceous solids; (5) Assessment of the potential of production of high yields of cycloalkanes by direct liquefaction of coal. Future high-Mach aircraft will place severe thermal demands on jet fuels, requiring the development of novel, hybrid fuel mixtures capable of withstanding temperatures in the range of 400--500 C. In the new aircraft, jet fuel will serve as both an energy source and a heat sink for cooling the airframe, engine, and system components. The ultimate development of such advanced fuels requires a thorough understanding of the thermal decomposition behavior of jet fuels under supercritical conditions. Considering that jet fuels consist of hundreds of compounds, this task must begin with a study of the thermal degradation behavior of select model compounds under supercritical conditions. The research performed by The Pennsylvania State University was focused on five major tasks that reflect the objectives stated above: Task 1: Investigation of the Quantitative Degradation of Fuels; Task 2: Investigation of Incipient Deposition; Task 3: Characterization of Solid Gums, Sediments, and Carbonaceous Deposits; Task 4: Coal-Based Fuel Stabilization Studies; and Task 5: Exploratory Studies on the Direct Conversion of Coal to High Quality Jet Fuels. The major findings of each of these tasks are presented in this executive summary. A description of the sub-tasks performed under each of these tasks and the findings of those studies are provided in the remainder of this volume

  10. Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures

    OpenAIRE

    Alessandro Petruzzi; Francesco D'Auria

    2008-01-01

    In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first system codes, developed at the beginning of the 1970s, utilized the homogenous equilibrium model with three balance equations to describe the two-phase flow, nowadays the more advanced system codes are based on the so-called “two-fluid model” with separation of the water and vapor phases, resulti...

  11. Regulatory requirements to the thermal-hydraulic and thermal-mechanical computer codes

    International Nuclear Information System (INIS)

    The paper presents an overview of the regulatory requirements to the thermal-hydraulic and thermal-mechanical computer codes, which are used for safety assessment of the fuel design and the fuel utilization. Some requirements to the model development, verification and validation of the codes and analysis of code uncertainties are also define. Questions concerning Quality Assurance during development and implementation of the codes as well as preparation of a detailed verification and validation plan are briefly discussed

  12. Thermal-hydraulic tests and analyses for the APR1400's development and licensing

    International Nuclear Information System (INIS)

    The program on Thermal-Hydraulic Evaluation by Testing and Analysis (THETA) for the development and licensing of the new design features in the APR1400 (Advanced Power Reactor-1400) is briefly introduced with a presentation on the research motivation and typical results of the separate effect tests and analyses of the major design features. The first part deals with multi-dimensional phenomena related to the safety analysis of the APR1400. One research area is related to the multi-dimensional behavior of the Safety Injection (SI) water in a reactor pressure vessel downcomer that uses a direct vessel injection type of SI system. The other area is associated with the condensation of steam jets and the resultant thermal mixing in a water pool; these phenomena are relevant to the depressurization of a Reactor Coolant System (RCS). The second part describes our efforts to develop new components for safety enhancements, such as a fluidic device as a passive SI flow controller and a sparger to depressurize the RCS. This work contributes to an understanding of the new thermal-hydraulic phenomena that are relevant to advanced reactor system designs; it also improves the prediction capabilities of analysis tools for multi-dimensional flow behavior, especially in complicated geometries

  13. Thermal hydraulic analysis of the AHWR—The Indian thorium fuelled innovative nuclear reactor

    International Nuclear Information System (INIS)

    Highlights: • Advanced heavy water reactor. • Thermal hydraulics. • Safety analysis. • RELAP5. -- Abstract: Analysis has been carried out for simulating loss-of-coolant accident (LOCA) at inlet header in a natural circulation type reactor developed as the advanced heavy water reactor (AHWR).The paper will cover a case of LOCA due to 200% break at inlet header which is double ended rupture. The maximum clad surface temperature has been predicted in different cases by using the thermal hydraulic safety code RELAP5/Mod4.0. The proposed reactor is a 920 MWth vertical pressure tube type, boiling light water cooled and heavy water moderated reactor. One of the important passive design features of this reactor is heat removal through natural circulation of primary coolant (at all allowed power levels) with no primary coolant pumps. This reactor is equipped with emergency core cooling system (ECCS) and isolation condensers (ICs) to remove decay heat during LOCA. This ECCS provides cooling to fuel in passive mode during first fifteen minutes of LOCA and it is achieved by high pressure injection from advanced accumulator. Cooling is continued for Later for three days by the gravity driven water pool (GDWP). This paper investigates the impact of high pressure injection in this cooling process

  14. Engineered Barrier Systems Thermal-Hydraulic-Chemical Column Test Report

    Energy Technology Data Exchange (ETDEWEB)

    W.E. Lowry

    2001-12-13

    The Engineered Barrier System (EBS) Thermal-Hydraulic-Chemical (THC) Column Tests provide data needed for model validation. The EBS Degradation, Flow, and Transport Process Modeling Report (PMR) will be based on supporting models for in-drift THC coupled processes, and the in-drift physical and chemical environment. These models describe the complex chemical interaction of EBS materials, including granular materials, with the thermal and hydrologic conditions that will be present in the repository emplacement drifts. Of particular interest are the coupled processes that result in mineral and salt dissolution/precipitation in the EBS environment. Test data are needed for thermal, hydrologic, and geochemical model validation and to support selection of introduced materials (CRWMS M&O 1999c). These column tests evaluated granular crushed tuff as potential invert ballast or backfill material, under accelerated thermal and hydrologic environments. The objectives of the THC column testing are to: (1) Characterize THC coupled processes that could affect performance of EBS components, particularly the magnitude of permeability reduction (increases or decreases), the nature of minerals produced, and chemical fractionation (i.e., concentrative separation of salts and minerals due to boiling-point elevation). (2) Generate data for validating THC predictive models that will support the EBS Degradation, Flow, and Transport PMR, Rev. 01.

  15. Engineered Barrier System Thermal-Hydraulic-Chemical Column Test Report

    International Nuclear Information System (INIS)

    The Engineered Barrier System (EBS) Thermal-Hydraulic-Chemical (THC) Column Tests provide data needed for model validation. The EBS Degradation, Flow, and Transport Process Modeling Report (PMR) will be based on supporting models for in-drift THC coupled processes, and the in-drift physical and chemical environment. These models describe the complex chemical interaction of EBS materials, including granular materials, with the thermal and hydrologic conditions that will be present in the repository emplacement drifts. Of particular interest are the coupled processes that result in mineral and salt dissolution/precipitation in the EBS environment. Test data are needed for thermal, hydrologic, and geochemical model validation and to support selection of introduced materials (CRWMS M and O 1999c). These column tests evaluated granular crushed tuff as potential invert ballast or backfill material, under accelerated thermal and hydrologic environments. The objectives of the THC column testing are to: (1) Characterize THC coupled processes that could affect performance of EBS components, particularly the magnitude of permeability reduction (increases or decreases), the nature of minerals produced, and chemical fractionation (i.e., concentrative separation of salts and minerals due to boiling-point elevation). (2) Generate data for validating THC predictive models that will support the EBS Degradation, Flow, and Transport PMR, Rev. 01

  16. ASME proceedings of the 32nd national heat transfer conference (HTD-Vol. 350). Volume 12: Fundamental experiment techniques in heat transfer; Thermal hydraulics of advanced nuclear reactors; Heat and mass transfer in supercritical liquid systems; Heat transfer in energy conversion; Heat transfer equipment; Heat transfer in gas turbine systems

    International Nuclear Information System (INIS)

    This volume contains a portion of the over 240 ASME papers which were presented at the conference. For over 40 years, the National Heat Transfer Conference has been the premiere forum for the presentation and dissemination of the latest advances in heat transfer. The work contained in these volumes range from studies of fundamental phenomena to applications in the latest heat transfer equipment. Topics covered in this volume are: Fundamental experiment techniques in heat transfer; thermal hydraulics of advanced nuclear reactors; heat and mass transfer in supercritical fluid systems; heat transfer in energy conversion; heat transfer equipment; and heat transfer in gas turbine systems. Separate abstracts were prepared for most papers in this volume

  17. Thermal hydraulic performance assessment of dual-cooled annular nuclear fuel for OPR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Chang-Hwan, E-mail: shinch@kaeri.re.kr [LWR Fuel Development Division, Korea Atomic Energy Research Institute, 989-111 Daedeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Chun, Tae-Hyun, E-mail: thchun@kaeri.re.kr [LWR Fuel Development Division, Korea Atomic Energy Research Institute, 989-111 Daedeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Oh, Dong-Seok, E-mail: dsoh1@kaeri.re.kr [LWR Fuel Development Division, Korea Atomic Energy Research Institute, 989-111 Daedeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); In, Wang-Kee, E-mail: wkin@kaeri.re.kr [LWR Fuel Development Division, Korea Atomic Energy Research Institute, 989-111 Daedeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of)

    2012-02-15

    Highlights: Black-Right-Pointing-Pointer A thermal hydraulic performance of a 12 Multiplication-Sign 12 annular fuel array is evaluated. Black-Right-Pointing-Pointer The subchannel analysis code for the dual-cooled annular fuel, MATRA-AF is validated. Black-Right-Pointing-Pointer We evaluate the sensitivity for geometry tolerances and operating parameter. Black-Right-Pointing-Pointer We decide the essential design parameters to uprate the power generation by dual-cooled annular fuel. Black-Right-Pointing-Pointer A thermal margin amount accommodating a 20% power-uprate seems viable. - Abstract: An internally and externally cooled annular fuel was proposed for an advance PWR, which can endure substantial power uprating. KAERI is pursuing the development for a reloading of power uprated annular fuel for the operating PWR reactors of OPR-1000. In this paper, the characteristics and verification of the MATRA-AF are described. The thermal hydraulic performance of a 12 Multiplication-Sign 12 annular fuel is calculated for the major design parameters and its performance is compared against the reference 16 Multiplication-Sign 16 cylindrical fuel assembly. In particular, the enhancements of the thermal hydraulic performance of dual-cooled annular fuel are estimated for the 100% normal power reactor core. The purpose of this study is to estimate a normal power for OPR-1000 with dual-cooled annular fuel, and ultimately to assess the feasibility of 120% core power. The parametric study was carried out for the fuel rod dimension, gap conductance, thermal diffusion coefficients, and pressure loss of the spacer grids. As a result of the analysis on the nominal power, annular fuel showed a sufficient margin available on DNB and fuel pellet temperature relative to cylindrical fuel. The margin amount seems accommodating a 20% power-uprate seems viable.

  18. Comparative Analysis of CTF and Trace Thermal-Hydraulic Codes Using OECD/NRC PSBT Benchmark Void Distribution Database

    Directory of Open Access Journals (Sweden)

    M. Avramova

    2013-01-01

    Full Text Available The international OECD/NRC PSBT benchmark has been established to provide a test bed for assessing the capabilities of thermal-hydraulic codes and to encourage advancement in the analysis of fluid flow in rod bundles. The benchmark was based on one of the most valuable databases identified for the thermal-hydraulics modeling developed by NUPEC, Japan. The database includes void fraction and departure from nucleate boiling measurements in a representative PWR fuel assembly. On behalf of the benchmark team, PSU in collaboration with US NRC has performed supporting calculations using the PSU in-house advanced thermal-hydraulic subchannel code CTF and the US NRC system code TRACE. CTF is a version of COBRA-TF whose models have been continuously improved and validated by the RDFMG group at PSU. TRACE is a reactor systems code developed by US NRC to analyze transient and steady-state thermal-hydraulic behavior in LWRs and it has been designed to perform best-estimate analyses of LOCA, operational transients, and other accident scenarios in PWRs and BWRs. The paper presents CTF and TRACE models for the PSBT void distribution exercises. Code-to-code and code-to-data comparisons are provided along with a discussion of the void generation and void distribution models available in the two codes.

  19. Thermal-hydraulic experiments and analyses on cold moderator

    International Nuclear Information System (INIS)

    A cold moderator using supercritical hydrogen is one of the key components in a MW-scale spallation target system, which directly affects the neutronic performance both in intensity and resolution. Since a hydrogen temperature rise in the moderator vessel affects the neutronic performance, it is necessary to suppress the local temperature rise within 3 K. In order to develop the conceptual design of the moderator structure in progress, the flow patterns were measured using a PIV (Particle Image Velocimeter) system under water flow conditions using a flat model that simulated a moderator vessel. From these results, the flow patterns (such as recirculation flows, stagnant flows etc.) were clarified. The hydraulic analytical results obtained using the STAR-CD code agreed well with experimental results. Thermal-hydraulic analyses in the moderator vessel were carried out using the STAR-CD code. Based on these results, we clarified the possibility of suppressing the local temperature rise to within 3 K under 2 MW operating conditions. In order to achieve the cost decreasing of the hydrogen loop, it is necessary to operate it reducing the hydrogen flow rate and the whole hydrogen mass. Then improved moderator concept using blowholes and a twisted tape was proposed, and we have tried to examine the effect of the blowing flow from the inlet pipe. From the experimental and analytical results, the blowing flow could be feasible for the suppression of the stagnant region. (author)

  20. Neutronics and thermal-hydraulics analyses of the pellet bed reactor for nuclear thermal propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Morley, N.J.; El-Genk, S. [Univ. of New Mexico, Albuquerque, NM (United States)

    1995-01-01

    Neutronics and thermal-hydraulics design and analyses of the pellet bed reactor for nuclear thermal propulsion are performed based on consideration of reactor criticality, passive decay heat removal, maximum fuel temperature, and subcriticality during a water flooding accident. Besides calculating the dimensions of the reactor core to satisfy the excess reactivity requirement at the beginning-of-mission of 1.25 $ (K{sub eff} of 1.01), the TWODANT discrete ordinates code is used to estimate the radial and axial fission power density profiles in the core. These power profiles are used in the nuclear propulsion thermal-hydraulic analysis model (NUTHAM-S) to determine the two-dimensional steady-state temperature, pressure, and flow fields in the core and optimize the orificing in the hot frit to avoid hot spots in the core at full-power operation.

  1. Neutronics and thermal-hydraulics analyses of the pellet bed reactor for nuclear thermal propulsion

    International Nuclear Information System (INIS)

    Neutronics and thermal-hydraulics design and analyses of the pellet bed reactor for nuclear thermal propulsion are performed based on consideration of reactor criticality, passive decay heat removal, maximum fuel temperature, and subcriticality during a water flooding accident. Besides calculating the dimensions of the reactor core to satisfy the excess reactivity requirement at the beginning-of-mission of 1.25 $ (Keff of 1.01), the TWODANT discrete ordinates code is used to estimate the radial and axial fission power density profiles in the core. These power profiles are used in the nuclear propulsion thermal-hydraulic analysis model (NUTHAM-S) to determine the two-dimensional steady-state temperature, pressure, and flow fields in the core and optimize the orificing in the hot frit to avoid hot spots in the core at full-power operation

  2. Resolution of thermal-hydraulic safety and licensing issues for the system 80+{sup {trademark}} design

    Energy Technology Data Exchange (ETDEWEB)

    Carpentino, S.E.; Ritterbusch, S.E.; Schneider, R.E. [ABB-Combustion Engineering, Windsor, CT (United States)] [and others

    1995-09-01

    The System 80+{sup {trademark}} Standard Design is an evolutionary Advanced Light Water Reactor (ALWR) with a generating capacity of 3931 MWt (1350 MWe). The Final Design Approval (FDA) for this design was issued by the Nuclear Regulatory Commission (NRC) in July 1994. The design certification by the NRC is anticipated by the end of 1995 or early 1996. NRC review of the System 80+ design has involved several new safety issues never before addressed in a regulatory atmosphere. In addition, conformance with the Electric Power Research Institute (EPRI) ALWR Utility Requirements Document (URD) required that the System 80+ plant address nuclear industry concerns with regard to design, construction, operation and maintenance of nuclear power plants. A large number of these issues/concerns deals with previously unresolved generic thermal-hydraulic safety issues and severe accident prevention and mitigation. This paper discusses the thermal-hydraulic analyses and evaluations performed for the System 80+ design to resolve safety and licensing issues relevant to both the Nuclear Stream Supply System (NSSS) and containment designs. For the NSSS design, the Safety Depressurization System mitigation capability and resolution of the boron dilution concern are described. Examples of containment design issues dealing with containment shell strength, robustness of the reactor cavity walls and hydrogen mixing under severe accident conditions are also provided. Finally, the overall approach used in the application of NRC`s new (NUREG-1465) radiological source term for System 80+ evaluation is described. The robustness of the System 80+ containment design to withstand severe accident consequences was demonstrated through detailed thermal-hydraulic analyses and evaluations. This advanced design to shown to meet NRC severe accident policy goals and ALWR URD requirements without any special design features and unnecessary costs.

  3. Relevant thermal-hydraulic aspects in the design of the RRR (Replacement Research Reactor)

    International Nuclear Information System (INIS)

    A description of the main thermal-hydraulic features and challenges of the Replacement Research Reactor, for the Australian Nuclear Science and Technology Organization (ANSTO), is presented. Different hydraulic and thermal-hydraulic aspects are considered, core cooling during full power operation and the way it affects the design, design criteria, engineered safety features and computational tools, amongst others. A special section is devoted to the thermal-hydraulic aspects inside the reflector tank, as well as the cooling of irradiation facilities, particularly, the Molybdenum production facility. (author)

  4. Thermal hydraulic analysis of the encapsulated nuclear heat source

    International Nuclear Information System (INIS)

    An analysis has been carried out of the steady state thermal hydraulic performance of the Encapsulated Nuclear Heat Source (ENHS) 125 MWt, heavy liquid metal coolant (HLMC) reactor concept at nominal operating power and shutdown decay heat levels. The analysis includes the development and application of correlation-type analytical solutions based upon first principles modeling of the ENHS concept that encompass both pure as well as gas injection augmented natural circulation conditions, and primary-to-intermediate coolant heat transfer. The results indicate that natural circulation of the primary coolant is effective in removing heat from the core and transferring it to the intermediate coolant without the attainment of excessive coolant temperatures. (authors)

  5. Core thermal hydraulic analysis for TNR power uprating

    International Nuclear Information System (INIS)

    This paper presents preliminary results of a study undertaken to investigate the possibility of raising the power of the Tajura Nuclear Research Reactor (TNRR) from 10 to 20 MWt keeping the same core configuration and with minimum changes in the primary cooling circuit. The study was carried out for a fresh core, with compact load (16 assemblies) under normal operation conditions. A computer program, TAJT, was used to simulate the core and perform the necessary thermal hydraulic analysis. The results obtained show that the reactor power could be raised to 15 MWt safely and with no changes in the primary cooling circuit. To raise the power to 20 MWt will require changes in the core configuration and primary circuit

  6. Hydraulic modeling of thermal discharges into shallow, tidal affected streams

    International Nuclear Information System (INIS)

    A two-unit nuclear fired power plant is being constructed in western Washington state. Blowdown water from cooling towers will be discharged into the Chehalis River nearby. The location of a diffuser is some 21 miles upriver from Grays Harbor on the Pacific Ocean. Because the Chehalis River is classified as an excellent stream from the standpoint of water quality, State regulatory agencies required demonstration that thermal discharges would maintain water quality standards within fairly strict limits. A hydraulic model investigation used a 1:12 scale, undistorted model of a 1300-foot river reach in the vicinity of the diffuser. The model scale was selected to insure fully turbulent flows both in the stream and from the diffuser (Reynolds similitude). Model operation followed the densimetric Froude similitude. Thermistors were employed to measure temperatures in the model; measurements were taken by computer command and such measurements at some 250 positions were effected in about 2.5 seconds

  7. Computer code for thermal hydraulic analysis of light water reactors

    International Nuclear Information System (INIS)

    A computer programme (THAL) has been developed to perform thermal hydraulic analysis of a single channel in a light water moderated core. In this code the hydrodynamic and thermodynamic equations describing one-dimensional axial flow have been discretized and solved explicitly stepwise up the coolant channel for an arbitrary power profile. THAL has been developed for use on small computers and it is capable of predicting the coolant, clad and fuel temperature profiles, steam quality, void fraction, pressure drop, critical heat flux and DNB ratio throughout the core. A boiling water reactor and a pressurized water reactor have been analyzed as test cases. The results obtained through the use of THAL compare favourably with those given in the design reports of these reactor systems. (author)

  8. Thermal-hydraulic analysis of the TR-2 reactor

    International Nuclear Information System (INIS)

    In this study the thermal-hydraulic analysis of the TR-2 reactor of 5 MW has been performed. The methods and results obtained are discussed. The methods which are used throughout the analysis can be applied to other plate type research reactors. For the present, the analysis are done for steady state conditions. For fuel elements and cooling channels the relations between pressure loss and flow rate, critical heat fluxes, safety margins and temperature distributions are calculated. The effects of UAlx-Al, U3O8-Al and U3Si2-Al type fuel materials on the peak fuel temperature are also studied. It has been found, assuming that the permissible minimum safety margin to onset of nucleate boiling be 2.32, the radial peaking factor should be lower than 3.5 and as far as the cooling system is unchanged this is also valid for low enriched fuels. (author)

  9. Analytical thermal hydraulic model for steam chugging phenomenon

    International Nuclear Information System (INIS)

    The Indian Pressurized Heavy Water Reactors (PHWRs) of the current design and Boiling Water Reactors incorporate a vapour suppression pool in the containment to mitigate the consequences of a loss of coolant accident. The thermal hydraulic phenomena occurring within the suppression pool following a loss of coolant accident are the vent clearing phase, pool swell and condensation of steam in the pool, which could be steady or intermittent depending on the steam mass flux. At low flow rates when steady condensation cannot be maintained at the steam-water interface, intermittent vapor condensation and bubble collapse process can be observed. This phenomenon is called steam chugging, during which rapid fluid and pressure oscillations occur. These oscillations may induce significant pressure loads on the submerged structures and are significant from safety perspective. This report addresses the problem of steam chugging. A thermal hydraulic model has been developed from fundamental conservation laws, for the process of steam chugging. The objective of developing the model is to present an approximation of the real phenomena and to obtain an analytical solution. The emphasis is however laid on studying the effect of presence of small amount of air in steam on chugging. Chugging is dominated by a number of important parameters hence, at the outset, a parametric study was undertaken using the above model to study the effect of important variables and to capture some essential features of the phenomena. This was done for a case when drywell contains only steam. Subsequently, the effect of presence of air in steam was studied using the non-condensable gas model. An attempt has been made to show numerically that the presence of a small amount of air in steam would effectively stabilize condensation and prevent inception chugging. Typical results are presented in this report. (author)

  10. Scaling in nuclear reactor system thermal-hydraulics

    International Nuclear Information System (INIS)

    Scaling is a reference 'key-word' in engineering and in physics. The relevance of scaling in the water cooled nuclear reactor technology constitutes the motivation for the present paper. The origin of the scaling-issue, i.e. the impossibility to get access to measured data in case of accident in nuclear reactors, is discussed at first. The so-called 'scaling-controversy' constitutes an outcome. Then, a critical survey (or 'scaling state-of-art';) is given of the attempts and of the approaches to provide a solution to the scaling-issue in the area of Nuclear Reactor System Thermal-Hydraulics (NRSTH): dimensionless design factors for Integral Test Facilities (ITF) are distinguished from scaling factors. The last part of the paper has a two-fold nature: (a) classifying the information about achievements in the area of thermal-hydraulics which are relevant to scaling: the concepts of 'scaling-pyramid' and the related 'scaling bridges' are introduced; (b) establishing a logical path across the scaling achievements (represented as a 'scaling puzzle'). In this context, the 'roadmap for scaling' is proposed: the objective is addressing the scaling issue when demonstrating the applicability of system codes in the licensing process of nuclear power plants. The code itself is referred hereafter as the 'key-to-scaling'. The database from the operation of properly scaled ITF and the availability of qualified system codes are identified as main achievements in NRSTH connected with scaling. The 'roadmap to scaling' constitutes a unified approach to scaling which aims at solving the 'scaling puzzle' created by researches performed during a half-a-century period.

  11. Scaling in nuclear reactor system thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    D' Auria, F., E-mail: dauria@ing.unipi.i [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, University of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy); Galassi, G.M. [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, University of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)

    2010-10-15

    Scaling is a reference 'key-word' in engineering and in physics. The relevance of scaling in the water cooled nuclear reactor technology constitutes the motivation for the present paper. The origin of the scaling-issue, i.e. the impossibility to get access to measured data in case of accident in nuclear reactors, is discussed at first. The so-called 'scaling-controversy' constitutes an outcome. Then, a critical survey (or 'scaling state-of-art';) is given of the attempts and of the approaches to provide a solution to the scaling-issue in the area of Nuclear Reactor System Thermal-Hydraulics (NRSTH): dimensionless design factors for Integral Test Facilities (ITF) are distinguished from scaling factors. The last part of the paper has a two-fold nature: (a) classifying the information about achievements in the area of thermal-hydraulics which are relevant to scaling: the concepts of 'scaling-pyramid' and the related 'scaling bridges' are introduced; (b) establishing a logical path across the scaling achievements (represented as a 'scaling puzzle'). In this context, the 'roadmap for scaling' is proposed: the objective is addressing the scaling issue when demonstrating the applicability of system codes in the licensing process of nuclear power plants. The code itself is referred hereafter as the 'key-to-scaling'. The database from the operation of properly scaled ITF and the availability of qualified system codes are identified as main achievements in NRSTH connected with scaling. The 'roadmap to scaling' constitutes a unified approach to scaling which aims at solving the 'scaling puzzle' created by researches performed during a half-a-century period.

  12. Simplified model of VVER–1000 thermal hydraulic process

    International Nuclear Information System (INIS)

    This report introduce developed mathematical model of thermal hydraulic process which occurs in the core of VVER – 1000 type of nuclear reactor and the coolant flow is considered in one dimension. Navier – Stokes differential equations system is taken like basis – namely continuity equation and momentum and energy conservation equations with two algebraic equations of state by which closure relationship is done. Following the approach of simplifying the model in momentum and energy equations some simplifying assumptions is made like Reynolds term in momentum equation is neglected and diffusion term – in energy equation. The differential equations system can be split into two parts: at one hand continuity equation and momentum equation are solved regarding velocity and pressure distribution and at another – solving energy equation. This report is considering the second case – solving the unsteady energy equation at prescribed distributions of velocity and pressure. By using one of the algebraic state equation, i=cvT, the energy equation is written regarding the temperature. Velocity and pressure given in the model are estimated by the thermal hydraulics means. The energy equation is solved by finite volumes method at which considered region is divided by N finite volumes, scalar values T and P are represented at central points at volumes and the velocity – at borders of these volumes by so called staggered grid. The equation is integrated at the boundaries of every finite volume and in time at the interval [t; t+ t]. For results obtained by integration is applied Crank-Nicolson semi-implicit scheme. As a result this gives an algebraic system with three diagonal matrix which can be solved with Crout effective algorithm.

  13. Preliminary thermal/hydraulic sizing calculations for duplex tube evaporator/superheater (interchangeable units). Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Waszink, R.P.; Hwang, J.Y.; Efferding, L.E.

    1974-06-01

    This is a preliminry thermal/hydraulic report reflecting work under Subtask 6.2 of Ref. 1.1. This report is an extension of the previous thermal/hydraulic design report. Parts of this report have been transmitted to GE. The detailed design basis, listed by source, is given. Additional details are discussed.

  14. Study on the thermal-hydraulic stability of high burn up STEP III fuel in Japan

    International Nuclear Information System (INIS)

    Japanese BWR utilities have performed a joint study of the Thermal Hydraulic Stability of High Burn up STEP III Fuel. In this study, the parametric dependency of thermal hydraulic stability threshold was obtained. It was confirmed through experiments that the STEP III Fuel has sufficient stability characteristics. (author)

  15. Development of Numerical Simulation Technology for High-Resolution Thermal Hydraulic Analysis

    International Nuclear Information System (INIS)

    A realistic simulation of two-phase flows is essential for the advanced design and safe operation of a nuclear reactor system. The need for a multi-dimensional analysis of thermal-hydraulics in nuclear reactor components is further increasing with advanced design features, such as a direct vessel injection system, a gravity-driven safety injection system, and a passive secondary cooling system. These features require more detailed analysis with enhanced accuracy. In this regard, KAERI has developed a three-dimensional thermal hydraulics code, CUPID, for the analysis of transient, multi-dimensional, two-phase flows in nuclear reactor components. The code was designed for use as a component-scale code, and/or a three-dimensional component, which can be coupled with a system code. This report presents an overview of the CUPID code development and preliminary assessment, mainly focusing on the numerical solution method and its verification and validation. It was shown that the CUPID code was successfully verified. The results of the validation calculations show that the CUPID code is very promising, but a systematic approach for the validation and improvement of the physical models is still needed

  16. Development of numerical simulation technology for high resolution thermal hydraulic analysis

    International Nuclear Information System (INIS)

    A realistic simulation of two phase flows is essential for the advanced design and safe operation of a nuclear reactor system. The need for a multi dimensional analysis of thermal hydraulics in nuclear reactor components is further increasing with advanced design features, such as a direct vessel injection system, a gravity driven safety injection system, and a passive secondary cooling system. These features require more detailed analysis with enhanced accuracy. In this regard, KAERI has developed a three dimensional thermal hydraulics code, CUPID, for the analysis of transient, multi dimensional, two phase flows in nuclear reactor components. The code was designed for use as a component scale code, and/or a three dimensional component, which can be coupled with a system code. This report presents an overview of the CUPID code development and preliminary assessment, mainly focusing on the numerical solution method and its verification and validation. It was shown that the CUPID code was successfully verified. The results of the validation calculations show that the CUPID code is very promising, but a systematic approach for the validation and improvement of the physical models is still needed

  17. CRISSUE-S, Neutronics/Thermal-hydraulics Coupling in LWR Technology

    International Nuclear Information System (INIS)

    Description: The CRISSUE-S project was created with the aim of re-evaluating fundamental technical issues in the technology of LWRs. Specifically, the project seeks to address the interactions between neutron kinetics and thermal-hydraulics that affect neutron moderation and influence the accident performance of the NPPs. This is undertaken in the light of the advanced computational tools that are readily available to the scientific community today. Specifically, the CRISSUE-S activity deals with the control of fission power and the use of high burn up fuel; these topics are part of the EC Work Programme as well as that of other international organisations such as the OECD/NEA and the IAEA. The problems of evaluating reactivity induced accident (RIA) consequences and eventually deciding the possibility of NPP prolongation must be addressed and resolved. RIA constitutes one of the most important of the ?less-resolved? safety issues, and treating this problem may have huge positive financial, social and environmental impacts. Public acceptance of nuclear technology implies that problems such as these be satisfactorily resolved. Cross-disciplinary (regulators, industry, utilities and research bodies) interaction and co operation within CRISSUE-S provides results which can directly and immediately be beneficial to EU industry. Co-operation at an international level: the participation of the EU, former Eastern European countries, the USA, and observers from Japan testify to the broad interest these problems engender. Competencies in broad areas such as thermal-hydraulics, neutronics and fuel, overall system design and reactor surveillance are needed to address the problems that are posed here. Excellent expertise is available in specific areas, while limited knowledge exists in the interface zones of those areas, e.g. in the coupling between thermal-hydraulics and neutronics. In general terms, the activities carried out and described here aim at exploiting available

  18. Verification of combined thermal-hydraulic and heat conduction analysis code FLOWNET/TRUMP

    International Nuclear Information System (INIS)

    This report presents the verification results of the combined thermal-hydraulic and heat conduction analysis code, FLOWNET/TRUMP which has been utilized for the core thermal hydraulic design, especially for the analysis of flow distribution among fuel block coolant channels, the determination of thermal boundary conditions for fuel block stress analysis and the estimation of fuel temperature in the case of fuel block coolant channel blockage accident in the design of the High Temperature Engineering Test Reactor(HTTR), which the Japan Atomic Energy Research Institute has been planning to construct in order to establish basic technologies for future advanced very high temperature gas-cooled reactors and to be served as an irradiation test reactor for promotion of innovative high temperature new frontier technologies. The verification of the code was done through the comparison between the analytical results and experimental results of the Helium Engineering Demonstration Loop Multi-channel Test Section(HENDEL T1-M) with simulated fuel rods and fuel blocks. (author)

  19. Monte Carlo and thermal-hydraulic coupling via PVMEXEC

    International Nuclear Information System (INIS)

    Successful high-fidelity coupling between a Monte Carlo neutron transport solver and a subchannel thermal-hydraulics solver has been achieved using PVMEXEC, a coupling frame-work developed for analysis of transient phenomenon in nuclear reactors. The PVMEXEC framework provides a generic program interface for exchanging data between solver kernels for different physical processes, such as radiation transport, heat conduction, and fluid flow. In this study, PVMEXEC was used to couple the in-house Monte Carlo radiation transport code, MC21, with a locally modified version of COBRA-TF. In this coupling scheme, MC21 is responsible for calculating three-dimensional power distributions and COBRA-TF for calculating local fluid temperatures and densities, as well as fuel temperatures. The coupled system was used to analyze 3D single-pin and assembly models based on the Calvert Cliffs commercial PWR. Convergence properties of the coupled simulations are examined and results are compared to simulations conducted using the existing integrated thermal feedback kernel in MC21. (author)

  20. Compatibility analysis of DUPIC fuel(4) - thermal hydraulic analysis

    International Nuclear Information System (INIS)

    Thermal-hydraulic compatibility of the DUPIC fuel bundle in the CANDU reactor has been studied. The critical channel power, the critical power ratio, the channel exit quality and the channel flow are calculated for the DUPIC and the standard fuels by using the NUCIRC code. The physical models and associated parametric values for the NUCIRC analysis of the fuels are also presented. Based upon the slave channel analysis, the critical channel power and the critical power ratios have been found to be very similar for the two fuel types. The same dryout model is used in this study for the standard and the DUPIC fuel bundles. To assess the dryout characteristics of the DUPIC fuel bundle, the ASSERT-PV code has been used for the subchannel analysis. Based upon the results of the subchannel analysis, it is found that the dryout location and the power for the two fuel types are indeed very similar. This study shows that thermal performance of the DUPIC fuel is not significantly different from that of the standard fuel

  1. Hydraulic performance of compacted clay liners under simulated daily thermal cycles.

    Science.gov (United States)

    Aldaeef, A A; Rayhani, M T

    2015-10-01

    Compacted clay liners (CCLs) are commonly used as hydraulic barriers in several landfill applications to isolate contaminants from the surrounding environment and minimize the escape of leachate from the landfill. Prior to waste placement in landfills, CCLs are often exposed to temperature fluctuations which can affect the hydraulic performance of the liner. Experimental research was carried out to evaluate the effects of daily thermal cycles on the hydraulic performance of CCLs under simulated landfill conditions. Hydraulic conductivity tests were conducted on different soil specimens after being exposed to various thermal and dehydration cycles. An increase in the CCL hydraulic conductivity of up to one order of magnitude was recorded after 30 thermal cycles for soils with low plasticity index (PI = 9.5%). However, medium (PI = 25%) and high (PI = 37.2%) plasticity soils did not show significant hydraulic deviation due to their self-healing potential. Overlaying the CCL with a cover layer minimized the effects of daily thermal cycles, and maintained stable hydraulic performance in the CCLs even after exposure to 60 thermal cycles. Wet-dry cycles had a significant impact on the hydraulic aspect of low plasticity CCLs. However, medium and high plasticity CCLs maintained constant hydraulic performance throughout the test intervals. The study underscores the importance of protecting the CCL from exposure to atmosphere through covering it by a layer of geomembrane or an interim soil layer. PMID:26241932

  2. Theoretical and experimental studies of heavy liquid metal thermal hydraulics. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    Through the Nuclear Energy Department's Technical Working Group on Fast Reactors (TWG-FR), the IAEA provides a forum for exchange of information on national programmes, collaborative assessments, knowledge preservation, and cooperative research in areas agreed by the Member States with fast reactor and partitioning and transmutation development programmes (e.g. accelerator driven systems (ADS)). Trends in advanced fast reactor and ADS designs and technology development are periodically summarized in status reports, symposia, and seminar proceedings prepared by the IAEA to provide all interested IAEA Member States with balanced and objective information. The use of heavy liquid metals (HLM) is rapidly diffusing in different research and industrial fields. The detailed knowledge of the basic thermal hydraulics phenomena associated with their use is a necessary step for the development of the numerical codes to be used in the engineering design of HLM components. This is particularly true in the case of lead or lead-bismuth eutectic alloy cooled fast reactors, high power particle beam targets and in the case of the cooling of accelerator driven sub-critical cores where the use of computational fluid dynamic (CFD) design codes is mandatory. Periodic information exchange within the frame of the TWG-FR has lead to the conclusion that the experience in HLM thermal fluid dynamics with regard to both the theoretical/numerical and experimental fields was limited and somehow dispersed. This is the case, e.g. when considering turbulent exchange phenomena, free-surface problems, and two-phase flows. Consequently, Member States representatives participating in the 35th Annual Meeting of the TWG-FR (Karlsruhe, Germany, 22-26 April 2002) recommended holding a technical meeting (TM) on Theoretical and Experimental Studies of Heavy Liquid Metal Thermal Hydraulics. Following this recommendation, the IAEA has convened the Technical Meeting on Theoretical and Experimental Studies of

  3. Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor

    International Nuclear Information System (INIS)

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R and D) that will be critical to the success of the NGNP, primarily in the areas of: (1) High temperature gas reactor fuels behavior; (2) High temperature materials qualification; (3) Design methods development and validation; (4) Hydrogen production technologies; and (5) Energy conversion. This paper presents current R and D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs

  4. Thermal hydraulics of the very high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R and D) that will be critical to the success of the NGNP, primarily in the areas of: · High temperature gas reactor fuels behavior · High temperature materials qualification · Design methods development and validation · Hydrogen production technologies · Energy conversion. This paper presents current R and D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs. (author)

  5. Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; Davie Petti

    2009-10-01

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: • High temperature gas reactor fuels behavior • High temperature materials qualification • Design methods development and validation • Hydrogen production technologies • Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  6. Application of integral mass and momentum balance method to simulation of CANDU thermal-hydraulic processes

    International Nuclear Information System (INIS)

    An integral modeling technique is presented, which converts a single-fluid or a two-fluid flow model into a low order dynamical system of ordinary differential equations. This method is applied to simulation of thermal-hydraulic CANDU transients. Some typical results will be reported with comparison to the predictions by using another modeling method. The results presented in this paper are obtained from simulation of a loss of coolant accident with a 100 % break of an outlet collector and with the reactor at nominal power level for a fuel channel situated on the intact loop of circuit. Some of the main results for the fuel channel outlet are presented. The integral balances of mass, momentum and energy over different types of control volumes result in a set of ordinary differential equations, which can be easily solved. Due to its flexibility and compatibility this modeling technique can be used as a basis to develop a full-range and unified modeling system to include both advantages of the best estimate code and a fast running code. This is particularly useful for an advanced plant analyzer or a full scope real time simulation to achieve both high-speed and detailed simulation of nuclear thermal-hydraulic phenomena. The test of the thermal-hydraulic model and resolution method performed until now confirmed that a global mass and a loop momentum balance can simulate adequately at least most transients with subcooled liquid and up to liquid dominant two-phase flow with low void fraction. (authors)

  7. Thermal Hydraulic Analysis Using GIS on Application of HTR to Thermal Recovery of Heavy Oil Reservoirs

    Directory of Open Access Journals (Sweden)

    Yangping Zhou

    2012-01-01

    Full Text Available At present, large water demand and carbon dioxide (CO2 emissions have emerged as challenges of steam injection for oil thermal recovery. This paper proposed a strategy of superheated steam injection by the high-temperature gas-cooled reactor (HTR for thermal recovery of heavy oil, which has less demand of water and emission of CO2. The paper outlines the problems of conventional steam injection and addresses the advantages of superheated steam injection by HTR from the aspects of technology, economy, and environment. A Geographic Information System (GIS embedded with a thermal hydraulic analysis function is designed and developed to analyze the strategy, which can make the analysis work more practical and credible. Thermal hydraulic analysis using this GIS is carried out by applying this strategy to a reference heavy oil field. Two kinds of injection are considered and compared: wet steam injection by conventional boilers and superheated steam injection by HTR. The heat loss, pressure drop, and possible phase transformation are calculated and analyzed when the steam flows through the pipeline and well tube and is finally injected into the oil reservoir. The result shows that the superheated steam injection from HTR is applicable and promising for thermal recovery of heavy oil reservoirs.

  8. 75 FR 80544 - NUREG-1953, Confirmatory Thermal-Hydraulic Analysis To Support Specific Success Criteria in the...

    Science.gov (United States)

    2010-12-22

    ... COMMISSION NUREG-1953, Confirmatory Thermal-Hydraulic Analysis To Support Specific Success Criteria in the..., ``Confirmatory Thermal-Hydraulic Analysis to Support Specific Success Criteria in the Standardized Plant Analysis... . SUPPLEMENTARY INFORMATION: NUREG-1953, ``Confirmatory Thermal-Hydraulic Analysis to Support Specific...

  9. Development of a frequency-domain coupled neutronic thermal-hydraulic stability analysis code STAC. Verification of thermal-hydraulic part of the code

    International Nuclear Information System (INIS)

    A frequency-domain coupled neutronic thermal-hydraulic stability analysis code STAC is under development in TEPCO Systems Corporation (TEPSYS). The code is composed of the steady-state thermal-hydraulic calculation part and the transfer function calculation part. In the transfer function calculation part, neutronics, fuel heat conduction and thermal-hydraulics models are implemented. In this paper, the thermal hydraulic part of the code is focused on. A basic framework of the code is learned from NUFREQ-NP code. The basic equations are almost the same, but many modifications are conducted. The major modifications in the thermal-hydraulic part are 1) introduction of the finite difference scheme in spatial discretization to simplify the code structure and facilitate the modification of the code and 2) consideration of perturbation in almost all empirical correlations. The STAC code is validated using steady-state pressure drop and void fraction data which were measured in NUPEC full bundle test. A good agreement between predicted and measured values is shown thus the steady-state calculation part of the code is well validated. Then, the STAC code is validated using stability threshold power measurement test data. The stability threshold power is calculated by STAC and the predicted and measured values are compared. Those values agree well. Predicted and measured resonance frequencies are also compared and good agreement is observed. (author)

  10. Thermal-hydraulic experiments for the PCHE type steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Shin, C. W.; No, H. C. [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    Printed circuit heat exchanger (PCHE) manufactured by HEATRIC is a compact type of the mini-channel heat exchanger. The PCHE is manufactured by diffusion bonding of the chemically-etched plates, and has high heat transfer rate due to a large surface. Therefore, the size of heat exchanger can be reduced by 1/5 - 1/6 and PCHE can be operated under high pressure, high temperature and multi-phase flow. Under such merits, it is used as heat exchanger with various purposes of gas cycle and water cycle. Recently, it is newly suggested as an application of a steam generator. IRIS of MIT and FASES of KAIST conceptually adopted PCHE as a steam generator. When using boiling condition of micro-channel, flow instability is one of the critical issues. Instability may cause unstable mass flow rate, sudden temperature change and system control failure. However instability tests of micro channels using water are very limited because the previous studies were focused on a single tube or other fluid instead of water. In KAIST, we construct the test facility to study the thermal hydraulics and fluid dynamics of the heat exchanger, especially occurrence of instability. By inducing the pressure drop of inlet water, amplitude of oscillation declined by 90%. Finally, the throttling effect was experimentally confirmed that PCHE could be utilized as a steam generator.

  11. Thermal Hydraulic Analysis Of Thorium-Based Annular Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Han, Kyu Hyun [Korea Institute of Nuclear Safety, 19, Guseong-dong, Yuseong-gu, Daejeon, 305-338 (Korea, Republic of)

    2008-07-01

    Thermal hydraulic characteristics of thorium-based fuel assemblies loaded with annular seed pins have been analyzed using AMAP combined with MATRA, and compared with those of the existing thorium-based assemblies. MATRA and AMAP showed good agreements for the pressure drops at the internal sub-channels. The pressure drop generally increased in the cases of the assemblies loaded with annular seed pins due to the larger wetted perimeter, but an exception existed. In the inner sub-channels of the seed pins, mass fluxes were high due to the grid form losses in the outer sub-channels. About 43% of the heat generated from the seed pin flowed into the inner sub-channel and the rest into the outer sub-channel, which implies the inner to outer wall heat flux ratio was approximately 1.2. The maximum temperatures of the annular seed pins were slightly above 500 deg. C. The MDNBRs of the assemblies loaded with annular seed pins were higher than those of the existing assemblies. Due to the fact that inter-channel mixing cannot occur in the inner sub-channels, temperatures and enthalpies were higher in the inner sub-channels. (author)

  12. Thermal Hydraulic Analysis Of Thorium-Based Annular Fuel Assemblies

    International Nuclear Information System (INIS)

    Thermal hydraulic characteristics of thorium-based fuel assemblies loaded with annular seed pins have been analyzed using AMAP combined with MATRA, and compared with those of the existing thorium-based assemblies. MATRA and AMAP showed good agreements for the pressure drops at the internal sub-channels. The pressure drop generally increased in the cases of the assemblies loaded with annular seed pins due to the larger wetted perimeter, but an exception existed. In the inner sub-channels of the seed pins, mass fluxes were high due to the grid form losses in the outer sub-channels. About 43% of the heat generated from the seed pin flowed into the inner sub-channel and the rest into the outer sub-channel, which implies the inner to outer wall heat flux ratio was approximately 1.2. The maximum temperatures of the annular seed pins were slightly above 500 deg. C. The MDNBRs of the assemblies loaded with annular seed pins were higher than those of the existing assemblies. Due to the fact that inter-channel mixing cannot occur in the inner sub-channels, temperatures and enthalpies were higher in the inner sub-channels. (author)

  13. Thermal hydraulic analysis of the JMTR improved LEU-core

    Energy Technology Data Exchange (ETDEWEB)

    Tabata, Toshio; Nagao, Yoshiharu; Komukai, Bunsaku; Naka, Michihiro; Fujiki, Kazuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Takeda, Takashi [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokai, Ibaraki (Japan)

    2003-01-01

    After the investigation of the new core arrangement for the JMTR reactor in order to enhance the fuel burn-up and consequently extend the operation period, the ''improved LEU core'' that utilized 2 additional fuel elements instead of formerly installed reflector elements, was adopted. This report describes the results of the thermal-hydraulic analysis of the improved LEU core as a part of safety analysis for the licensing. The analysis covers steady state, abnormal operational transients and accidents, which were described in the annexes of the licensing documents as design bases events. Calculation conditions for the computer codes were conservatively determined based on the neutronic analysis results and others. The results of the analysis, that revealed the safety criteria were satisfied on the fuel temperature, DNBR and primary coolant temperature, were used in the licensing. The operation license of the JMTR with the improved LEU core was granted in March 2001, and the reactor operation with new core started in November 2001 as 142nd operation cycle. (author)

  14. Basis for Coupled 3-D Neutronics and Thermal-Hydraulics

    International Nuclear Information System (INIS)

    The purpose of this seminar is first to discuss the basis of the coupling between 3-D Neutron- Kinetics and Thermal-Hydraulics codes, including the control and 3-D variables to interchange, the transform of the 3-D NK and TH core nodalizations, and the schemes for temporal coupling and time-step control. As representative examples of the NK-TH core coupling, we discuss first the integration of a 3-D NK nodal code with a TH subchannel code, for detailed transient core analysis; and second the coupling of 3-D NK nodal codes with TH system codes, for general transient and safety analysis. In chapter 2, we analyze several prototype model transients in PWR, where large 3-D core asymmetries are found and the NK-TH coupling is quite significant, including loss-of-flow and symmetric and asymmetric core cooling, considering the effects on the responses of the excore detectors. In chapter 3, we discuss the analysis of an increase-of-flow transient actually occurred in an operating PWR and the comparison with the measured data. In chapter 4, we summarize the phenomena and results of the calculations of the NEA/NSC Benchmark on the main steam line break (MSLB) transient in a PWR. Finally, we will discuss the state-of-the-art issues in LWR coupled NK-TH 3-D transient analysis and ongoing and planned computational developments.

  15. Sensitivity theory applied to a transient thermal-hydraulics problem

    Energy Technology Data Exchange (ETDEWEB)

    Weber, C.F.; Oblow, E.M.

    1979-10-01

    A new method for sensitivity analysis of transient nonlinear problems is developed and applied to a reactor thermal-hydraulics problem. The method resembles the differential sensitivity methods currently used in the linear problems of reactor physics, but it is applicable to nonlinear systems as well. The equations governing heat transfer and fluid flow in a fuel pin and surrounding coolant are given and used to derive a second set of equations (commonly known as the adjoint equations) used in the sensitivity analysis. Both systems contain one second-order parabolic and one first-order hyperbolic partial differential equation. Difference equations are derived to approximate both systems and the convergence properties of these discrete systems are evaluated, yielding a useful analysis of the numerical solution. The solution functions are used to derive sensitivity coefficients for any desired integral response. These sensitivity coefficients are used in a first-order perturbation theory to predict changes in a response resulting from changes in parameter values. The results of a test problem are shown, verifying that this procedure is indeed useful for a wide variety of sensitivity calculations.

  16. Thermal hydraulics of accelerator driven system windowless targets

    Directory of Open Access Journals (Sweden)

    Bruno ePanella

    2015-07-01

    Full Text Available The study of the fluid dynamics of the windowless spallation target of an Accelerator Driven System (ADS is presented. Several target mockup configurations have been investigated: the first one was a symmetrical target, that was made by two concentric cylinders, the other configurations are not symmetrical. In the experiments water has been used as hydraulic equivalent to lead-bismuth eutectic fluid. The experiments have been carried out at room temperature and flow rate up to 24 kg/s. The fluid velocity components have been measured by an ultrasound technique. The velocity field of the liquid within the target region either for the approximately axial-symmetrical configuration or for the not symmetrical ones as a function of the flow rate and the initial liquid level is presented. A comparison of experimental data with the prediction of the finite volume FLUENT code is also presented. Moreover the results of a 2D-3D numerical analysis that investigates the effect on the steady state thermal and flow fields due to the insertion of guide vanes in the windowless target unit of the EFIT project ADS nuclear reactor are presented, by analysing both the cold flow case (absence of power generation and the hot flow case (nominal power generation inside the target unit.

  17. Development of fuel performance and thermal hydraulic technology

    International Nuclear Information System (INIS)

    Space grid in LWR fuel assembly is a key structural component to support fuel rods and to enhance heat transfer from fuel rod to the coolant. Therefore, the original spacer grid has been developed. In addition, new phenomena in fuel behavior occurs at the high burnup, so that models to analyze those new phenomena were developed. Results of this project can be summarized as follows. - Seven different spacer grid candidates have been invented and submitted for domestic and US patents. Spacer grid test specimen(3x3 array and 5x5 array) were fabricated for each candidate and the mechanical tests were performed. - Basic technologies in the mechanical and thermal hydraulic behavior in the spacer grid development are studied and relevant test facilities were established - Fuel performance analysis models and programs were developed for the high burnup pellet and cladding, and fuel performance data base were compiled - Procedures of fuel characterization and in-/out of-pile tests were prepared - Conceptual design of fuel rod for integral PWR was carried out. (author)

  18. Nuclear reactor thermal hydraulics safety analysis and thoughts on FUKUSHIMA

    International Nuclear Information System (INIS)

    The first part of this article is to show my thoughts on the accident at Fukushima Daiichi Nuclear Power Station. It is cited from a summary of my lecture talk in Indonesia, in the beginning of the last December, 2011. This talk was based on my previous lecture and seminar talks including those delivered at MIT, June 16, at the ANS Annual Meeting in Hollywood, Florida, June 28 at NURETH-13 in Toronto, September 27, and others. The content is based on the open and latest information available to date in Japan. It may contain some erroneous or uncertain information. I tried to minimize it to my best capability. Also I tried to eliminate any critical issues or opinions that may jeopardize some people who were involved in. The latter half of this article will be excerpts of my recent R and D activities related to the safety-by-design for sodium cooled fast reactors and light water reactors, thermal hydraulics analysis focusing on the simulation-based technology, in particular subchannel analysis and computational fluid dynamics. (J.P.N.)

  19. Development of fuel performance and thermal hydraulic technology

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Youn Ho; Song, K. N.; Kim, H. K. and others

    2000-03-01

    Space grid in LWR fuel assembly is a key structural component to support fuel rods and to enhance heat transfer from fuel rod to the coolant. Therefore, the original spacer grid has been developed. In addition, new phenomena in fuel behavior occurs at the high burnup, so that models to analyze those new phenomena were developed. Results of this project can be summarized as follows. - Seven different spacer grid candidates have been invented and submitted for domestic and US patents. Spacer grid test specimen(3x3 array and 5x5 array) were fabricated for each candidate and the mechanical tests were performed. - Basic technologies in the mechanical and thermal hydraulic behavior in the spacer grid development are studied and relevant test facilities were established - Fuel performance analysis models and programs were developed for the high burnup pellet and cladding, and fuel performance data base were compiled - Procedures of fuel characterization and in-/out of-pile tests were prepared - Conceptual design of fuel rod for integral PWR was carried out. (author)

  20. Sub Channel Thermal hydraulics Design Analysis of PWR-KSNP

    International Nuclear Information System (INIS)

    Sub channel analysis for the fuel element of thermal hydraulics design PWR-KSNP reactor has been carried out. PWR-KSNP reactor is a kind of Pressurized Water Reactor (PWR) Nuclear Power Plant developed by Korea (Korean Standard Nuclear Plant), that produce an electricity power about 1000 MWe. In the analysis, a fuel assembly with 4 fuel rods piled up into matrix 2 x 2, and surrounding by 9 sub channels of coolant, was used as a calculation model. There are 3 models of fuel assembly, i.e. the radial factors in the first model are 1.144, 1.144, 1.120 and 1.121, in the second fuel model are 0.994 , 1.005 , 0.987 and 0.989, and in the third model are 2.500, 1.144, 1.120 and 1.121, respectively. The calculated results using the COBRA IV-I code showed that the maximum cladding temperature revolved by 340.3 - 349.0 ℃, the maximum temperature of meat surface (outer of meat) revolved by 498.1 - 758.2 ℃ and the maximum temperature of meat center revolved by 928.5 - 1843.7 ℃, respectively. Whereas the safety margin against DNBR revolved by 6.50 - 2.05. By maximum meat temperature limit of 2804 ℃ and the minimum DNBR of 1.30, it is concluded that the PWR-KSNP design was in the range of safety. (author)

  1. Thermal-hydraulic experiments for the PCHE type steam generator

    International Nuclear Information System (INIS)

    Printed circuit heat exchanger (PCHE) manufactured by HEATRIC is a compact type of the mini-channel heat exchanger. The PCHE is manufactured by diffusion bonding of the chemically-etched plates, and has high heat transfer rate due to a large surface. Therefore, the size of heat exchanger can be reduced by 1/5 - 1/6 and PCHE can be operated under high pressure, high temperature and multi-phase flow. Under such merits, it is used as heat exchanger with various purposes of gas cycle and water cycle. Recently, it is newly suggested as an application of a steam generator. IRIS of MIT and FASES of KAIST conceptually adopted PCHE as a steam generator. When using boiling condition of micro-channel, flow instability is one of the critical issues. Instability may cause unstable mass flow rate, sudden temperature change and system control failure. However instability tests of micro channels using water are very limited because the previous studies were focused on a single tube or other fluid instead of water. In KAIST, we construct the test facility to study the thermal hydraulics and fluid dynamics of the heat exchanger, especially occurrence of instability. By inducing the pressure drop of inlet water, amplitude of oscillation declined by 90%. Finally, the throttling effect was experimentally confirmed that PCHE could be utilized as a steam generator

  2. One-dimensional two-phase thermal hydraulics (ENSTA course)

    International Nuclear Information System (INIS)

    This course is part of the ENSTA 3rd year thermal hydraulics program (nuclear power option). Its purpose is to provide the theoretical basis and main physical notions pertaining to two-phase flow, mainly focussed on water-steam flows. The introduction describes the physical specificities of these flows, emphasizing their complexity. The mathematical bases are then presented (partial derivative equations), leading to a one-dimensional type, simplified description. Balances drawn up for a pipe length volume are used to introduce the mass conservation. motion and energy equations for each phase. Various postulates used to simplify two-phase models are presented, culminating in homogeneous model definitions and equations, several common examples of which are given. The model is then applied to the calculation of pressure drops in two-phase flows. This involves presenting the models most frequently used to represent pressure drops by friction or due to pipe irregularities, without giving details (numerical values of parameters). This chapter terminates with a brief description of static and dynamic instabilities in two-phase flows. Finally, heat transfer conditions frequently encountered in liquid-steam flows are described, still in the context of a 1D model. This chapter notably includes reference to under-saturated boiling conditions and the various forms of DNB. The empirical heat transfer laws are not discussed in detail. Additional material is appended, some of which is in the form of corrected exercises. (author). 6 appends

  3. Thermal-hydraulic safety analysis for CH-HCSB TBM

    International Nuclear Information System (INIS)

    CH-HCSB TBM is designed to be tested in ITER by Southwestern Institute of Physics and its aim is to validate the feasibility of DEMO fusion reactor. The thermal-hydraulic safety analysis has to testify that the TBM and its Helium Cooling System (HCS) will not impact the safe operation of ITER under accidental conditions. In order to simulate the transient accidents, TBM and HCS are modeled using RELAP5. The steady state results indicate that the designed TBM input/output temperatures are obtained and the temperature of FW Beryllium armor is limited to the reasonable range. The Ex-Vessel LOCA is very dangerous because of the melting of FW Beryllium armor after about 80 seconds of the happened LOCA and some controlling measures have to be taken before melting. It's not so dangerous for the In-Vessel LOCA and In-Box LOCA, but the Tritium Extraction System has to be cut off from the HCS quickly when In-Box LOCA happens. Based on the results, the design of CH-HCSB TBM will be modified in order to assure the safety of TBM and ITER. (orig.)

  4. Coupling of FRAPTRAN Fuel Rod for Transient Analysis with GENFLO Thermal Hydraulic Code (KOTO and READY)

    International Nuclear Information System (INIS)

    Reactor analyses are becoming more and more challenging. Due to pursuing higher fuel discharge burnups, fuel designs and operational conditions are subject to constant upgrading. Advances in hardware have removed many of the limitations on detail of analyses, and best-estimate type applications have become a commonplace, a practice now increasingly adopted even in safety cases. At the same time there is a regulatory trend extending the range of events the licensee is to consider. Neutronics, thermal hydraulics, and fuel behaviour are closely interlinked during a reactor transient and cannot be generally separated in a realistic description. However, attempts to combine these into coupled models have been, even at their best, impracticably heavy to use. At VTT, in consent with the USNRC, an in-house general flow model GENFLO has now been coupled with the NRC's FRAPTRAN fuel performance code. The combination takes benefits of a fast-running non-iterative thermal hydraulic model and an updated fuel performance code validated for burnups of up to 65 MWd/kgU. A code description and results of two types of analyses are given. One is a hypothetical large break loss-of-coolant accident (LBLOCA) in a VVER reactor, the other is an instability transient in a Boiling Water Reactor (BWR), for which system conditions were separately available. A systematic validation and international peer review will follow. (orig.)

  5. FLICA-OVAP: Elements of validation for LWRs thermal-hydraulic studies

    International Nuclear Information System (INIS)

    FLICA-OVAP is an advanced two-phase flow thermal-hydraulics code based on a full 3-Dimensional subchannel approach. It is designed to analyze flows in Light Water Reactors (LWRs) cores such as PWRs, BWRs and experimental reactors. Therefore its applicability covers all ranges of operating conditions for water-cooled reactors. The FLICA-OVAP code includes several models, to adapt to the needs associated to different core concepts and multiple industrial and research applications. The set of models includes a Homogeneous Equilibrium Model (HEM), a four-equation drift flux model, a two-fluid model and a more general multi-field model. Several correlations are available to account for momentum, heat and mass transfer phenomena, as well as turbulence effects. This paper presents an overview of FLICA-OVAP modelling capabilities for applications in nuclear reactors design and safety analysis. A validation matrix is proposed and its results are presented. The matrix covers a wide range of selected phenomena, which are relevant for thermal-hydraulics studies. Therefore the different FLICA-OVAP physical correlations addressed in the current study include single phase and two-phase friction factors, single phase and boiling heat transfer, turbulence and critical heat flux. Results of the FLICA-OVAP validation studies highlight the capabilities of the code to well-predict two-phase flows in Light Water Reactors for both normal operation and under accidental circumstances. Future developments as well as validation activities are also summarized. (authors)

  6. Thermal Hydraulic design parameters study for severe accidents using neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Chang Hyun; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Chang, Keun Sun [Sunmoon University, Asan (Korea, Republic of)

    1997-12-31

    To provide the information on severe accident progression is very important for advanced or new type of nuclear power plant (NPP) design. A parametric study, therefore, was performed to investigate the effect of thermal hydraulic design parameters on severe accident progression of pressurized water reactors (PWRs). Nine parameters, which are considered important in NPP design or severe accident progression, were selected among the various thermal hydraulic design parameters. The backpropagation neural network (BPN) was used to determine parameters, which might more strongly affect the severe accident progression, among nine parameters. For training, different input patterns were generated by the latin hypercube sampling (LHS) technique and then different target patterns that contain core uncovery time and vessel failure time were obtained for Young Gwang Nuclear (YGN) Units 3 and 4 using modular accident analysis program (MAAP) 3.0B code. Three different severe accident scenarios, such as two loss of coolant accidents (LOCAs) and station blackout (SBO), were considered in this analysis. Results indicated that design parameters related to refueling water storage tank (RWST), accumulator and steam generator (S/G) have more dominant effects on the progression of severe accidents investigated, compared to the other six parameters. 9 refs., 5 tabs. (Author)

  7. Three-dimensional thermal-hydraulic response in LBLOCA based on MARS-KS calculation

    International Nuclear Information System (INIS)

    Three-dimensional (3D) thermal-hydraulic analysis of an accident in Nuclear Power Plant (NPP) has been extended to use since Best-Estimate (BE) calculation was allowed for safety analysis. The present study is to discuss why and how big differences can be obtained from the 1D and 3D thermal-hydraulic calculations for large break Loss-of-Coolant Accident (LBLOCA). Calculations are performed using MARS-KS code with one-dimensional (1D) modeling and with 3D modeling for reactor vessel of Advanced Power Reactor (APR1400). For the 3D modeling, the MULTI-D component of the MARS-KS code is applied. Especially, a hot channel having a size of one fuel assembly is also simulated. From the comparison of the calculation results, four differences are found: lower blowdown Peak Cladding Temperature (PCT) in 3D calculation, instantaneous stop of cladding heat-up, extent of blowdown quenching, and milder and longer reflood process in 3D calculation. The flow distribution in the core in 3D calculation could be one of the reasons for those differences. From the sensitivity study, the initial temperature at the reactor vessel upper head is found to have strong effect on the blowdown quenching, thus the reflood PCT and needs a careful consideration. (author)

  8. Evaluation of hot spot factors for thermal and hydraulic design of HTTR

    International Nuclear Information System (INIS)

    High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal power and 950degC in reactor outlet coolant temperature. One of the major items in thermal and hydraulic design of the HTTR is to evaluate the maximum fuel temperature with a sufficient margin from a viewpoint of integrity of coated fuel particles. Hot spot factors are considered in the thermal and hydraulic design to evaluate the fuel temperature not only under the normal operation condition but also under any transient condition conservatively. This report summarizes the items of hot spot factors selected in the thermal and hydraulic design and their estimated values, and also presents evaluation results of the thermal and hydraulic characteristics of the HTTR briefly. (author)

  9. Thermal-hydraulic transient analysis of a packed particle bed reactor fuel element

    OpenAIRE

    Casey, William Emerson

    1990-01-01

    Title as it appears in the M.I.T. Graduate List, Jun. 4, 1990: Transient thermal-hydraulic analysis of a packed particle bed reactor fuel element A model which describes the thermal-hydraulic behavior of a packed particle bed reactor fuel element is developed and compared to a reference standard. The model represents a step toward a thermal-hydraulic module for a real-time, autonomous reactor powder controller. The general configuration of the fuel element is a bed of small (diameter about...

  10. HERA: Hydro Engineering Reactor Applications for Thermal Hydraulics Operational Safety

    International Nuclear Information System (INIS)

    Nuclear energy continues to be faced with major challenges such as waste, economics, proliferation and last, but not least, safety. The Three Mile Island Unit 2 (TMI-2) and Chernobyl Unit 4 accidents have underscored the importance of safety design and operation to the future use of nuclear power. In particular, the in-vessel retention (IVR) is one of major severe accident management strategies adopted by a number of operating nuclear power plants (NPPs) during a severe accident. If there is inadequate cooling during a severe accident, a significant amount of core material could become molten and relocate to the lower head of the reactor pressure vessel (RPV) as was the case in the TMI-2 accident. Should it be possible to ensure that the RPV shall remain intact so that the relocated materials are retained within the RPV, the enhanced safety associated with these NPPs can possibly reduce concerns about the source term. The two APR1400s under construction as Shin-Gori Units 3 and 4 in the Republic of Korea adopted the external reactor vessel cooling (ERVC) by way of reactor cavity flooding as one of the major severe accident management strategies. The ERVC in the APR1400 design applies the active flooding through the thermal insulator. However, it is not clear whether the currently proposed ERVC without an additional apparatus could provide with sufficient heat removal for successful IVR in advanced and higher power reactors like APR+ (up to 1600 MWe). HERA (Hydro Engineering Reactor Applications) calls for cutting-edge technologies resorting to AIRIS (Arranged Intellectual Reactor Integral System), BASIS (Boiling Advanced Safety Integral System), and OASIS (Operational Advanced Safety Integral System) together with PARIS (Prototype Advanced Reactor Instrumentation System). HERA is aimed at developing engineering solutions to cope with transients and accidents in a coherent, continual, comprehensive manner. AIRIS clings to FILA (Flooding Integrated Layout Arrangement

  11. Thermal Margin Budget of a Simplified Core Thermal-Hydraulic Code for OPR1000

    International Nuclear Information System (INIS)

    A thermal-hydraulic analysis of a pressurized water reactor (PWR) core is usually conducted by a subchannel analysis method to prove a safe and reliable operation of a reactor. The reactor core is divided into a number of subchannels within which the thermal-hydraulic conditions are considered to be radially uniform. The coolant moves through the subchannels formed between neighboring fuel rods and between the peripheral fuel rods and the reactor core shroud. A subchannel code such as THINC-IV and TORC solves the mass, momentum and energy equations for the subchannels by the finite-difference method. They calculate the minimum departure from nucleate boiling ratio (DNBR) in a PWR core which is a measure for the core thermal margin. A simplified thermal-hydraulic code, CETOP-D, was developed to quickly calculate the minimum DNBR (MDNBR) based on a four-channel core model. A three-dimensional transport coefficient model is used to radially group a flow subchannel into a 4-channel core representation. The CETOP-D model also includes an adjusted hot assembly inlet flow factor to account for the deviations in the MDNBR due to a code simplification. The hot assembly flow factor is adjusted to eliminate a possible non-conservatism in the MDNBR prediction by the CETOP-D code. The CETOP-D code and its simplified versions are used to calculate the MDNBR for on-line core monitoring and protection systems as well as a safety analysis for a Korea optimized PWR, OPR1000. The purpose of this study is to estimate the conservatism in CETOP-D MDNBR and the potential DNBR margin enhancement for the CETOP-D applications. The MDNBR values by the TORC and CETOP-D codes were compared for a wide range of operating conditions for the OPR1000

  12. 78 FR 8202 - Meeting of the Joint ACRS Subcommittees on Thermal Hydraulic Phenomena and Materials, Metallurgy...

    Science.gov (United States)

    2013-02-05

    ... ACRS meetings were published in the Federal Register on October 18, 2012, (77 FR 64146- 64147... Hydraulic Phenomena and Materials, Metallurgy and Reactor Fuels; Notice of Meeting The Joint ACRS Subcommittees on Thermal Hydraulic Phenomena and Materials, Metallurgy and Reactor Fuels will hold a meeting...

  13. Thermal-hydraulics of lead bismuth for accelerator driven systems

    International Nuclear Information System (INIS)

    Full text of publication follows: Lead bismuth has been selected as one of the most suitable coolants to be used in accelerator driven systems (ADS) for transmutation of minor actinides. It serves both, as a target material of the spallation source to balance the neutron economy, and as a coolant with high thermal inertia to provide a safe and reliable heat transfer to the secondary power cycle. With the aim to develop the required technologies to enable the later design of such ADS systems, the Karlsruhe Lead bismuth LAboratory KALLA, consisting of three test loops, has been built and set into operation at the Forschungszentrum Karlsruhe since 2000, keeping more than 45 t of PbBi in operation at temperatures up to 550 deg. C. The test program includes oxygen control systems, heat flux simulation tools, electro-magnetic and mechanical pump technologies, heat transfer and flow measurements, reliability and corrosion tests. In a first test campaign, a technology loop called THESYS was built to develop measurement technologies for the acquisition of scalar quantities, like pressures, temperatures, concentrations, and flow rates, as well as velocity fields, which are required for both operational and scientific purposes. THESYS also allowed to perform generic turbulent heat transfer experiments necessary to provide liquid metal adapted turbulent heat transfer models for ADS design analyses. The second loop, the thermalhydraulic loop THEADES with an installed power of 2.5 MW, has been built to conduct prototypical component experiments for beam windows (e.g. MEGAPIE or MYHRRA) or fuel rod configurations. First test results will be reported. The experimental team is supported by a numerical team who studied the thermal hydraulics of the tested components in order to enable a later transfer of the results to industrial systems. Three different types of codes are being improved: lumped parameter codes (e.g. ATHLET) to perform system analyses for lead bismuth in loops

  14. FY 1995 progress report on the ANS thermal-hydraulic test loop operation and results

    International Nuclear Information System (INIS)

    The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. Special consideration was given to allow operation of the system in a stiff mode (constant flow) and in a soft mode (constant pressure drop) for proper implementation of true FE and DNB experiments. The facility is also designed to examine other T/H phenomena, including onset of incipient boiling (IB), single-phase heat transfer coefficients and friction factors, and two-phase heat transfer and pressure drop characteristics. Tests will also be conducted that are representative of decay heat levels at both high pressure and low pressure as well as other quasi-equilibrium conditions encountered during transient scenarios. A total of 22 FE tests and 2 CHF tests were performed during FY 1994 and FY 1995 with water flowing vertically upward. Comparison of these data as well as extensive data from other investigators led to a proposed modification to the Saha and Zuber correlation for onset of significant void (OSV), applied to FE prediction. The modification takes into account a demonstrated dependence of the OSV or FE thermal limits on subcooling levels, especially in the low subcooling regime

  15. Development of thermal hydraulic models for the reliable regulatory auditing code

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Song, C. H.; Lee, Y. J.; Kwon, T. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2003-04-15

    The objective of this project is to develop thermal hydraulic models for use in improving the reliability of the regulatory auditing codes. The current year fall under the first step of the 3 year project, and the main researches were focused on identifying the candidate thermal hydraulic models for improvement and to develop prototypical model development. During the current year, the verification calculations submitted for the APR 1400 design certification have been reviewed, the experimental data from the MIDAS DVI experiment facility in KAERI have been analyzed and evaluated, candidate thermal hydraulic models for improvement have been identified, prototypical models for the improved thermal hydraulic models have been developed, items for experiment in connection with the model development have been identified, and preliminary design of the experiment has been carried out.

  16. Methodology of thermal hydraulic analysis for substantiation of reactor vessel brittle fracture resistance

    International Nuclear Information System (INIS)

    Methodology of thermal hydraulic analysis for substantiation of reactor vessel brittle fracture resistance is presented in this article. This procedure was used during PTS study for SUNPP Unit 1 and represents generally accepted international approach.

  17. Thermal-hydraulic analysis of the pellet bed reactor for nuclear thermal propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Morley, N.J. (Institute for Space Nuclear Power Studies, Department of Chemical and Nuclear Engineering, University of New Mexico, Albuquerque, NM 87131-1341 (United States)); El-Genk, M.S. (Institute for Space Nuclear Power Studies, Department of Chemical and Nuclear Engineering, University of New Mexico, Albuquerque, NM 87131-1341 (United States))

    1994-09-01

    A two-dimensional steady-state thermal-hydraulics analysis of the pellet bed reactor for nuclear thermal propulsion is performed using the NUTHAM- S thermal-hydraulic code. The effects of axial heat and momentum transfers on the temperature and flow fields in the core are investigated. In addition, the porosity profile in the hot frit is optimized to avoid the development of a hot spot in the reactor core. Finally, a sensitivity analysis is performed using the optimized hot frit porosity profile to determine the effects of varying the propellant and core parameters on the peak fuel temperature and pressure drop across the core. These parameters include the inlet temperature and mass flow rate of the hydrogen propellant, average porosity of the core bed, the porosity of the hot frit, and local hot frit blockage. The peak temperature of the fuel is shown not to exceed its melting point as a result of changing any of these parameters from the base case, with the exception of hot frit blockage greater than 60% over a 0.12m axial segment of the hot frit. ((orig.))

  18. Scientific Design of Large Scale Sodium Thermal-Hydraulic Test Facility in KAERI

    International Nuclear Information System (INIS)

    A full passive decay heat removal system is implemented as an advanced design feature for the SFR which is currently being developed in Korea. Its operation depends purely on the natural circulation in a primary heat transport system and a passive decay heat removal system, and no active component or operator action is required. For the demonstration of the design concept, a large scale sodium thermal-hydraulic test facility is being designed with the plan of installation in 2013. In the experiments, the cooling capability during the long- and short-term periods after reactor shutdown will be demonstrated and also the produced experimental data will be utilized for the assessment and verification of the safety and performance analysis codes. In this paper, the preliminary design features of the test facility are presented along with the design requirements and methodology. (author)

  19. Nuclear thermal-hydraulics education: the Yankee Atomic/University of Lowell experience

    International Nuclear Information System (INIS)

    This paper summarizes the long and meaningful relationship between the University of Lowell (UL) and Yankee Atomic Electric Company (YAEC) in the area of nuclear thermal hydraulics. The UL has actively interacted with YAEC for many years. Many UL graduates from the nuclear program as well as health physics and other disciplines are employed by YAEC. Furthermore, many students have worked for YAEC on a part-time basis through summer employment or the coop program. Several graduate students have completed their thesis work under the joint direction of UL and YAEC personnel, and some faculty members have had consulting and research contracts with the company. At the same time, YAEC employees have taken advantage of the graduate program offered by UL and have earned advanced degrees. Some YAEC personnel have taught courses at UL and have served on the industrial advisory committees

  20. Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled molten salt reactors

    Energy Technology Data Exchange (ETDEWEB)

    Qiu, Suizheng; Zhang, Dalin; Liu, Minghao; Liu, Limin; Xu, Rongshuan; Gong, Cheng; Su, Guanghui [Xi' an Jiaotong Univ. (China). State Key Laboratory of Multiphase Flow in Power Engineering

    2016-05-15

    Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.

  1. Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled molten salt reactors

    International Nuclear Information System (INIS)

    Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.

  2. The status of studies on fast reactor core thermal hydraulics at PNC

    International Nuclear Information System (INIS)

    An outlook was addressed on investigative activities of the fast reactor core thermal-hydraulics at Power Reactor and Nuclear Fuel Development Corporation. Firstly, a computational modeling to predict flow field under natural circulation decay heat removal condition using multi-dimensional codes and its validation were presented. The validation was carried out through calculations of sodium experiments on an inter-subassembly heat transfer, a transient from forced to natural circulation and an inter-wrapper flow. Secondly, experimental and computational studies were expressed on local blockage with porous media in a fuel subassembly. Lastly, information was presented on an advanced computational code based on a subchannel analysis code. The code is under the development and extended to perform whole core simulation. (author)

  3. Analysis of thermal-hydraulics of a marine reactor in an oscillating acceleration field

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Hak; Park, Goon Cherl [Seoul National Univ., Seoul (Korea, Republic of)

    1996-07-01

    In this study the RETRAN-03 code was modified to analyze the thermal-hydraulic transients under three-dimensional ship motions for the application to the future marine reactors. First Japanese nuclear ship MUTSU reactor have been analyzed under various ship motions to verify this code. As results, typical thermal-hydraulic characteristics of marine reactors such as flow rate oscillations and S/G water level oscillations are successfully simulated at various conditions.

  4. Description of a stable scheme for steady-state coupled Monte Carlo-thermal-hydraulic calculations

    OpenAIRE

    Dufek, Jan; Eduard Hoogenboom, J.

    2014-01-01

    We provide a detailed description of a numerically stable and efficient coupling scheme for steady-state Monte Carlo neutronic calculations with thermal-hydraulic feedback. While we have previously derived and published the stochastic approximation based method for coupling the Monte Carlo criticality and thermal-hydraulic calculations, its possible implementation has not been described in a step-by-step manner. As the simple description of the coupling scheme was repeatedly requested from us...

  5. Steady-State Thermal-Hydraulic Analysis of TRIGA Research Reactor

    OpenAIRE

    Mohammad Mizanur Rahman; Mohammad Abdur R. Akond; Mohammad Khairul Basher; Md. Quamrul Huda

    2014-01-01

    The COOLOD-N2 and PARET computer codes were used for a steady-state thermal hydraulic and safety analysis of the 3 MW TRIGA Mark-II research reactor located at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh. The objective of the present study is to ensure that all important safety related thermal hydraulic parameters uphold margins far below the safety limits by steady-state calculations at full power. We, therefore, have calculated the hot channel fuel centreline ...

  6. Thermal Hydraulic Analysis of 3 MW TRIGA Research Reactor of Bangladesh Considering Different Cycles of Burnup

    OpenAIRE

    M. H. Altaf; N.H. Badrun

    2014-01-01

    Burnup dependent steady state thermal hydraulic analysis of TRIGA Mark-II research reactor has been carried out utilizing coupled point kinetics, neutronics and thermal hydraulics code EUREKA-2/RR. From the previous calculations of neutronics parameters including percentage burnup of individual fuel elements performed so far for 700 MWD burnt core of TRIGA reactor showed that the fuel rod predicted as hottest at the beginning of cycle (fresh core) was found to remain as the hottest until 200 ...

  7. Thermal-hydraulic Optimization of Water-cooled Center Conductor Post for Spherical Tokamaks Reactor

    Institute of Scientific and Technical Information of China (English)

    柯严; 吴宜灿; 黄群英; 郑善良

    2002-01-01

    This paper proposes a conceptual structure of segmental water-cooled Center Con ductor Post (CCP) to be flexible in installment and replacement. Thermal-hydraulic optimization and sensitivity analysis of key parameters are performed based on a reference fusion transmutation system with 100 MW fusion power. Numerical simulation by using a commercial code PHOEN]CS has been carried out to be close to the thermal-hydraulic analytical results of the CCP mid-part.

  8. Thermal-hydraulic code selection for modular high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    In order to study the transient thermal-hydraulic system behaviour of modular high temperature gas-cooled reactors, the thermal-hydraulic computer codes RELAP5, MELCOR, THATCH, MORECA, and VSOP are considered at the Netherlands Energy Research Foundation ECN. This report presents the selection of the most appropriate codes. To cover the range of relevant accidents, a suite of three codes is recommended for analyses of HTR-M and MHTGR reactors. (orig.)

  9. Current and anticipated uses of thermal-hydraulic codes in NFI

    Energy Technology Data Exchange (ETDEWEB)

    Tsuda, K. [Nuclear Fuel Industries, Ltd., Tokyo (Japan); Takayasu, M. [Nuclear Fuel Industries, Ltd., Sennann-gun (Japan)

    1997-07-01

    This paper presents the thermal-hydraulic codes currently used in NFI for the LWR fuel development and licensing application including transient and design basis accident analyses of LWR plants. The current status of the codes are described in the context of code capability, modeling feature, and experience of code application related to the fuel development and licensing. Finally, the anticipated use of the future thermal-hydraulic code in NFI is briefly given.

  10. Development of a preliminary PIRT (Phenomena Identification and Ranking Table) of thermal-hydraulic phenomena for SMART

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Lee, Won Jae; Kim, Hee Cheol; Song, Jin Ho; Sim, Suk Ku [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The work reported in this paper identifies the thermal-hydraulic phenomena that are expected to occur during a number of key transients in SMART (System-integrated Modular Advanced ReacTor) which is under development at KAERI. The result of this effort is based on the current design concept of SMART integral reactor. Although the design is still evolving, the preliminary Phenomena Identification and Ranking Table (PIRT) has been developed based on the experts` knowledge and experience. The preliminary PIRT has been developed by consensus of KAERI expert panelists and AHP (Analytical Hierarchy Process). Preliminary PIRT developed in this paper is intended to be used to identify and integrate development areas of further experimental tests needed, thermal hydraulic models and correlations and code improvements for the safety analysis of the SMART. 8 refs., 4 tabs (Author)

  11. Experimental thermal hydraulic facility for simulating LOCA behaviour of pressurised heavy water power reactor

    International Nuclear Information System (INIS)

    Experimental thermal hydraulic facility being set up adjacent to R and D Centre at Tarapur is a 13 MW full-elevation scaled down facility having the key components of PHT System of Pressurised Heavy Water Reactor (PHWR). The objective of the facility is to study thermal hydraulic behaviour of PHT System of PHWR by simulating various transients and accidental scenarios, to conduct safety related and operational transient studies and validation of various thermal hydraulic computer codes developed for analysis. The design of thermal hydraulic facility is based on the process parameters of a large PHWR with respect to fluid mass flux, transit time, flow velocity, pressure, temperature and enthalpy in PHT System. Experiments would be conducted in the facility to gain an improved understanding of the thermal hydraulic behaviour of large size PHWR during loss of coolant accident scenarios with forced and natural thermo-siphoning circulation modes etc. The data collected from the experiments would be used in validating computer codes developed for safety analysis. The facility is extensively instrumented to measure parameters such as temperature, pressure, flow, level, void-fraction at key locations. This paper gives the design philosophy used for scaling, design of major components of primary and secondary circuit of Experimental Thermal Hydraulic Facility and details of simulated experiments to be carried out. (author)

  12. FY 1993 progress report on the ANS thermal-hydraulic test loop operation and results

    Energy Technology Data Exchange (ETDEWEB)

    Siman-Tov, M.; Felde, D.K.; Farquharson, G. [and others

    1994-07-01

    The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). Highly subcooled heavy-water coolant flows vertically upward at a very high mass flux of almost 27 MG/m{sup 2}-s. In a parallel fuel plate configuration as in the ANSR, the flow is subject to a potential excursive static-flow instability that can very rapidly lead to flow starvation and departure from nucleate boiling (DNB) in the ``hot channel``. The current correlations and experimental data bases for flow excursion (FE) and critical heat flux (CHF) seldom evaluate the specific combination of ANSR operating parameters. The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. A series of FE tests with water flowing vertically upward was completed over a nominal heat flux range of 6 to 17 MW/m{sup 2}, a mass flux range of 8 to 28 Mg/m{sup 2}-s, an exit pressure range of 1.4 to 2.1 MPa, and an inlet temperature range of 40 to 50 C. FE experiments were also conducted using as ``soft`` a system as possible to secure a true FE phenomena (actual secondary burnout). True DNB experiments under similar conditions were also conducted. To the author`s knowledge, no other FE data have been reported in the literature to date that dover such a combination of conditions of high mass flux, high heat flux, and moderately high pressure.

  13. Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

    Energy Technology Data Exchange (ETDEWEB)

    Coryell, E.W.; Siefken, L.J.; Harvego, E.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States)] [and others

    1997-07-01

    The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures. The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.

  14. Thermal-Hydraulic Performance of Cross-Shaped Spiral Fuel in High-Power-Density BWRs

    International Nuclear Information System (INIS)

    Power up-rating of existing nuclear reactors promises to be an area of great study for years to come. One of the major approaches to efficiently increasing power density is by way of advanced fuel design, and cross-shaped spiral-fuel has shown such potential in previous studies. Our work aims to model the thermal-hydraulic consequences of filling a BWR core with these spiral-shaped pins. The helically-wound pins have a cross-section resembling a 4-petaled flower. They fill an assembly in a tight bundle, their dimensions chosen carefully such that the petals of neighboring pins contact each other at their outer-most extent in a self-supporting lattice, absent of grid spacers. Potential advantages of this design raise much optimism from a thermal-hydraulic perspective. These spiral rods possess about 40% larger surface area than traditional rods, resulting in increased cooling and a proportional reduction in average surface heat flux. The thin petal-like extensions help by lowering thermal resistance between the hot central region of the pin and the bulk coolant flow, decreasing the maximum fuel temperature by 200 deg. C according to Finite Element (COSMOS) models. However, COSMOS models also predict a potential problem area at the 'elbow' region of two adjoining petals, where heat flux peaking is twice that along the extensions. Preliminary VIPRE models, which account only for the surface area increase, predict a 22% increase in critical power. It is also anticipated that the spiral twist would provide the flowing coolant with an additional radial velocity component, and likely promote turbulence and mixing within an assembly. These factors are expected to provide further margin for increased power density, and are currently being incorporated into the VIPRE model. The reduction in pressure drop inherent in any core without grid-spacers is also expected to be significant in aiding core stability, though this has not yet been quantified. Spiral-fuel seems to be a

  15. Validation and Calibration of Nuclear Thermal Hydraulics Multiscale Multiphysics Models - Subcooled Flow Boiling Study

    Energy Technology Data Exchange (ETDEWEB)

    Anh Bui; Nam Dinh; Brian Williams

    2013-09-01

    In addition to validation data plan, development of advanced techniques for calibration and validation of complex multiscale, multiphysics nuclear reactor simulation codes are a main objective of the CASL VUQ plan. Advanced modeling of LWR systems normally involves a range of physico-chemical models describing multiple interacting phenomena, such as thermal hydraulics, reactor physics, coolant chemistry, etc., which occur over a wide range of spatial and temporal scales. To a large extent, the accuracy of (and uncertainty in) overall model predictions is determined by the correctness of various sub-models, which are not conservation-laws based, but empirically derived from measurement data. Such sub-models normally require extensive calibration before the models can be applied to analysis of real reactor problems. This work demonstrates a case study of calibration of a common model of subcooled flow boiling, which is an important multiscale, multiphysics phenomenon in LWR thermal hydraulics. The calibration process is based on a new strategy of model-data integration, in which, all sub-models are simultaneously analyzed and calibrated using multiple sets of data of different types. Specifically, both data on large-scale distributions of void fraction and fluid temperature and data on small-scale physics of wall evaporation were simultaneously used in this work’s calibration. In a departure from traditional (or common-sense) practice of tuning/calibrating complex models, a modern calibration technique based on statistical modeling and Bayesian inference was employed, which allowed simultaneous calibration of multiple sub-models (and related parameters) using different datasets. Quality of data (relevancy, scalability, and uncertainty) could be taken into consideration in the calibration process. This work presents a step forward in the development and realization of the “CIPS Validation Data Plan” at the Consortium for Advanced Simulation of LWRs to enable

  16. Development of Mitsubishi high thermal performance grid. CFD applicability for thermal hydraulic design

    International Nuclear Information System (INIS)

    Mitsubishi has developed a new zircalloy grid spacer for PWR fuel with higher thermal performance. Computational Fluid Dynamics (CFD) evaluation method has been applied for designing of the new lower pressure loss and higher Departure from Nucleate Boiling (DNB) benefit grid spacer. Reduction of pressure loss of grid structures has been examined by CFD. Also, CFD has been developed as a design tool to predict the coolant mixing ability of vane structures, which is to compare the relative peak spot temperatures around fuel rods at the same heat flux condition. Prototype grids were manufactured and several tests, which were pressure loss measurements, cross-flow measurements and freon DNB tests, were conducted to verify CFD predictions. It is concluded that the applicability of the CFD evaluation method for the thermal hydraulic design of the grid is confirmed. (author)

  17. Best-estimate plus uncertainty thermal-hydraulic stability analysis of BWRs using TRACG code

    International Nuclear Information System (INIS)

    on the Minimum Critical Power Ratio (MCPR) performance. The purpose of the DSS-CD TRACG analysis is to confirm the inherent MCPR margin afforded by the solution design. This paper presents the Best Estimate Plus Uncertainty (BEPU) DSS-CD TRACG methodology and its application to BWR Thermal-Hydraulic (T-H) stability analyses. The statistical Code Scaling, Applicability and Uncertainty (CSAU) methodology (defined in NUREG/CR-5249) is used to calculate the MCPR uncertainty. The TRACG simulation includes a full core individual bundle model in which each fuel bundle is modeled as an individual T-H channel. The complete CSAU analysis of full core individual bundle model is an innovative solution represents the state-of-the-art stability analysis of BWRs and is the first ever full statistical analysis for stability safety analyses. The adoption of BEPU methodologies for stability analyses advances the understanding of the associated physical phenomena and maintains the safety of reactor plant operation in expanded operation domain with up-rated power. (authors)

  18. Advanced hydraulic fracturing methods to create in situ reactive barriers

    International Nuclear Information System (INIS)

    This article describes the use of hydraulic fracturing to increase permeability in geologic formations where in-situ remedial action of contaminant plumes will be performed. Several in-situ treatment strategies are discussed including the use of hydraulic fracturing to create in situ redox zones for treatment of organics and inorganics. Hydraulic fracturing methods offer a mechanism for the in-situ treatment of gently dipping layers of reactive compounds. Specialized methods using real-time monitoring and a high-energy jet during fracturing allow the form of the fracture to be influenced, such as creation of assymmetric fractures beneath potential sources (i.e. tanks, pits, buildings) that should not be penetrated by boring. Some examples of field applications of this technique such as creating fractures filled with zero-valent iron to reductively dechlorinate halogenated hydrocarbons, and the use of granular activated carbon to adsorb compounds are discussed

  19. Thermal-Hydraulic Design of a Fluoride High-Temperature Demonstration Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, Juan J [ORNL; Qualls, A L [ORNL

    2016-01-01

    INTRODUCTION The Fluoride High-Temperature Reactor (FHR) named the Demonstration Reactor (DR) is a novel reactor concept using molten salt coolant and TRIstructural ISOtropic (TRISO) fuel that is being developed at Oak Ridge National Laboratory (ORNL). The objective of the FHR DR is to advance the technology readiness level of FHRs. The FHR DR will demonstrate technologies needed to close remaining gaps to commercial viability. The FHR DR has a thermal power of 100 MWt, very similar to the SmAHTR, another FHR ORNL concept (Refs. 1 and 2) with a power of 125 MWt. The FHR DR is also a small version of the Advanced High Temperature Reactor (AHTR), with a power of 3400 MWt, cooled by a molten salt and also being developed at ORNL (Ref. 3). The FHR DR combines three existing technologies: (1) high-temperature, low-pressure molten salt coolant, (2) high-temperature coated-particle TRISO fuel, (3) and passive decay heat cooling systems by using Direct Reactor Auxiliary Cooling Systems (DRACS). This paper presents FHR DR thermal-hydraulic design calculations.

  20. Determination of thermal-hydraulic loads on reactor internals in a DBA-situation

    International Nuclear Information System (INIS)

    Full text of publication follows: According to Finnish regulatory requirements, reactor internals have to stay intact in a design basis accident (DBA) situation, so that control rods can still penetrate into the core. To fulfill this demand some criteria must be followed in periodical in-service inspections. This is the motivation for studying and developing more detailed methods for analysis of thermal-hydraulic loads on reactor internals during the DBA-situation for the Loviisa NPP in Finland. The objective of this research program is to connect thermal-hydraulic and mechanical analysis methods with the goal to produce a reliable method for determination of thermal-hydraulic and mechanical loads on reactor internals in the accident situation. The tools studied are thermal-hydraulic system codes, computational fluid dynamics (CFD) codes and finite element analysis (FEA) codes. This paper concentrates mainly on thermal-hydraulic part of the research, but also the mechanical aspects are discussed. Firstly, the paper includes a short literary review of the available methods to analyse the described problem including both thermal-hydraulic and structural analysis parts. Secondly, different possibilities to carry out thermal-hydraulic analyses have been studied. The DBA-case includes complex physical phenomena and therefore modelling is difficult. The accident situation can be for example LLOCA. When the pipe has broken, the pressure decreases and water starts to evaporate, which consumes energy and that way limits the pressure decrease. After some period of time, the system reaches a new equilibrium state. To perform exact thermal-hydraulic analysis also two phase phenomena must be included. Therefore CFD codes are not capable of modelling the DBA situation very well, but the use of CFD codes requires that the effect of two phase flow must be added somehow. One method to calculate two phase phenomena with CFD codes is to use thermal-hydraulic system codes to calculate

  1. Determination of thermal-hydraulic loads on reactor internals in a DBA-situation

    Energy Technology Data Exchange (ETDEWEB)

    Ville Lestinen; Timo Toppila [POB 10, 00048 FORTUM (Finland)

    2005-07-01

    Full text of publication follows: According to Finnish regulatory requirements, reactor internals have to stay intact in a design basis accident (DBA) situation, so that control rods can still penetrate into the core. To fulfill this demand some criteria must be followed in periodical in-service inspections. This is the motivation for studying and developing more detailed methods for analysis of thermal-hydraulic loads on reactor internals during the DBA-situation for the Loviisa NPP in Finland. The objective of this research program is to connect thermal-hydraulic and mechanical analysis methods with the goal to produce a reliable method for determination of thermal-hydraulic and mechanical loads on reactor internals in the accident situation. The tools studied are thermal-hydraulic system codes, computational fluid dynamics (CFD) codes and finite element analysis (FEA) codes. This paper concentrates mainly on thermal-hydraulic part of the research, but also the mechanical aspects are discussed. Firstly, the paper includes a short literary review of the available methods to analyse the described problem including both thermal-hydraulic and structural analysis parts. Secondly, different possibilities to carry out thermal-hydraulic analyses have been studied. The DBA-case includes complex physical phenomena and therefore modelling is difficult. The accident situation can be for example LLOCA. When the pipe has broken, the pressure decreases and water starts to evaporate, which consumes energy and that way limits the pressure decrease. After some period of time, the system reaches a new equilibrium state. To perform exact thermal-hydraulic analysis also two phase phenomena must be included. Therefore CFD codes are not capable of modelling the DBA situation very well, but the use of CFD codes requires that the effect of two phase flow must be added somehow. One method to calculate two phase phenomena with CFD codes is to use thermal-hydraulic system codes to calculate

  2. Thermal hydraulics modeling of the US Geological Survey TRIGA reactor

    Science.gov (United States)

    Alkaabi, Ahmed K.

    The Geological Survey TRIGA reactor (GSTR) is a 1 MW Mark I TRIGA reactor located in Lakewood, Colorado. Single channel GSTR thermal hydraulics models built using RELAP5/MOD3.3, RELAP5-3D, TRACE, and COMSOL Multiphysics predict the fuel, outer clad, and coolant temperatures as a function of position in the core. The results from the RELAP5/MOD3.3, RELAP5-3D, and COMSOL models are similar. The TRACE model predicts significantly higher temperatures, potentially resulting from inappropriate convection correlations. To more accurately study the complex fluid flow patterns within the core, this research develops detailed RELAP5/MOD3.3 and COMSOL multichannel models of the GSTR core. The multichannel models predict lower fuel, outer clad, and coolant temperatures compared to the single channel models by up to 16.7°C, 4.8°C, and 9.6°C, respectively, as a result of the higher mass flow rates predicted by these models. The single channel models and the RELAP5/MOD3.3 multichannel model predict that the coolant temperatures in all fuel rings rise axially with core height, as the coolant in these models flows predominantly in the axial direction. The coolant temperatures predicted by the COMSOL multichannel model rise with core height in the B-, C-, and D-rings and peak and then decrease in the E-, F-, and G-rings, as the coolant tends to flow from the bottom sides of the core to the center of the core in this model. Experiments at the GSTR measured coolant temperatures in the GSTR core to validate the developed models. The axial temperature profiles measured in the GSTR show that the flow patterns predicted by the COMSOL multichannel model are consistent with the actual conditions in the core. Adjusting the RELAP5/MOD3.3 single and multichannel models by modifying the axial and cross-flow areas allow them to better predict the GSTR coolant temperatures; however, the adjusted models still fail to predict accurate axial temperature profiles in the E-, F-, and G-rings.

  3. Advanced Spacecraft Thermal Modeling Project

    Data.gov (United States)

    National Aeronautics and Space Administration — For spacecraft developers who spend millions to billions of dollars per unit and require 3 to 7 years to deploy, the LoadPath reduced-order (RO) modeling thermal...

  4. JPL Advanced Thermal Control Technology Roadmap - 2012

    Science.gov (United States)

    Birur, Gaj; Rodriguez, Jose I.

    2012-01-01

    NASA's new emphasis on human exploration program for missions beyond LEO requires development of innovative and revolutionary technologies. Thermal control requirements of future NASA science instruments and missions are very challenging and require advanced thermal control technologies. Limited resources requires organizations to cooperate and collaborate; government, industry, universities all need to work together for the successful development of these technologies.

  5. Experimental thermal hydraulic studies on the enhancement of safety od LWRs

    International Nuclear Information System (INIS)

    The safe use of nuclear power plants (NPPs) requires a deep understanding of the functioning of physical processes and systems involved. Studies on thermal hydraulics have been carried out in various separate effects and integral test facilities at Lappeenranta University of Technology (LUT) either to ensure the functioning of safety systems of light water reactors (LWR) or to produce validation data for the computer codes used in safety analyses of NPPs. Several examples of safety studies on thermal hydraulics of the nuclear power plants are discussed. Studies are related to the physical phenomena existing in different processes in NPPs, such as rewetting of the fuel rods, emergency core cooling (ECC), natural circulation, small break loss-of-coolant accidents (SBLOCA), non-condensable gas release and transport, and passive safety systems. Studies on both VVER and advanced light water reactor (ALWR) systems are included. The set of cases include separate effects tests for understanding and modeling a single physical phenomenon, separate effects tests to study the behavior of a NPP component or a single system, and integral tests to study the behavior of the whole system. In the studies following steps can be found, not necessarily in the same study. Experimental studies as such have provided solutions to existing design problems. Experimental data have been created to validate a single model in a computer code. Validated models are used in various transient analyses of scaled facilities or NPPs. Integral test data are used to validate the computer codes as whole, to see how the implemented models work together in a code. In the final stage test results from the facilities are transferred to the NPP scale using computer codes. Some of the experiments have confirmed the expected behavior of the system or procedure to be studied; in some experiments there have been certain unexpected phenomena that have caused changes to the original design to avoid the recognized

  6. Introducing and validating a new method for coupling neutronic and thermal-hydraulic calculations

    Energy Technology Data Exchange (ETDEWEB)

    Zare, Nafiseh [Department of Nuclear Engineering, Faculty of Engineering, Azad Islamic University, Science and Research- Branch, Punak Square, Tehran (Iran, Islamic Republic of); Fadaei, Amir Hosein, E-mail: Fadaei_amir@aut.ac.i [Faculty of Nuclear Engineering and Physics, Amirkabir University of Technology (Tehran Polytechnique), Hafez Street, Tehran (Iran, Islamic Republic of); Rahgoshay, Mohammad [Department of Nuclear Engineering, Faculty of Engineering, Azad Islamic University, Science and Research- Branch, Punak Square, Tehran (Iran, Islamic Republic of); Fadaei, Mohammad Mehdi [Department of Electrical Engineering, Faculty of Engineering, Central Tehran Branch, Islamic Azad University, Punak Square, Tehran (Iran, Islamic Republic of); Kia, Shabnam [Department of Nuclear Engineering, Faculty of Engineering, Azad Islamic University, Science and Research- Branch, Punak Square, Tehran (Iran, Islamic Republic of)

    2010-11-15

    Research highlights: {yields} Reactor behavior affects from reciprocal effects between neutronic and thermo-hydraulic. {yields} Reliable reactor analysis requires coupling of neutronic and thermal-hydraulic calculation. {yields} Iterative process can be used to perform neutronic and thermal-hydraulic calculation. - Abstract: In this study, a new and innovative method is introduced for analyzing neutronic and thermal-hydraulic calculation. For this aim, VVR-S research reactor was selected, and the calculation procedure was performed for it. WIMS, CITATION and COBRA-EN codes were used for investigation. Calculation model consists of two sub-models: neutronic and thermo-hydraulic. The neutronic model uses WIMS and CITATION codes for neutronic simulation of the reactor core and calculating flux and power distribution over it. WIMS code simulates the fuel assemblies and CITATION models the core. The thermal-hydraulic model uses COBRA-EN code for performing the relative calculation. In this study, FORTRAN 90 program is used for linking two sub-models and performing the calculation. The proposed procedure is performed for VVR-S analysis and finally, the obtained results are compared with the experimental results that show good agreement with it.

  7. Introducing and validating a new method for coupling neutronic and thermal-hydraulic calculations

    International Nuclear Information System (INIS)

    Research highlights: → Reactor behavior affects from reciprocal effects between neutronic and thermo-hydraulic. → Reliable reactor analysis requires coupling of neutronic and thermal-hydraulic calculation. → Iterative process can be used to perform neutronic and thermal-hydraulic calculation. - Abstract: In this study, a new and innovative method is introduced for analyzing neutronic and thermal-hydraulic calculation. For this aim, VVR-S research reactor was selected, and the calculation procedure was performed for it. WIMS, CITATION and COBRA-EN codes were used for investigation. Calculation model consists of two sub-models: neutronic and thermo-hydraulic. The neutronic model uses WIMS and CITATION codes for neutronic simulation of the reactor core and calculating flux and power distribution over it. WIMS code simulates the fuel assemblies and CITATION models the core. The thermal-hydraulic model uses COBRA-EN code for performing the relative calculation. In this study, FORTRAN 90 program is used for linking two sub-models and performing the calculation. The proposed procedure is performed for VVR-S analysis and finally, the obtained results are compared with the experimental results that show good agreement with it.

  8. Comparison for the interfacial and wall friction models in thermal-hydraulic system analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Moon Kyu; Park, Jee Won; Chung, Bub Dong; Kim, Soo Hyung; Kim, See Dal

    2007-07-15

    The average equations employed in the current thermal hydraulic analysis codes need to be closed with the appropriate models and correlations to specify the interphase phenomena along with fluid/structure interactions. This includes both thermal and mechanical interactions. Among the closure laws, an interfacial and wall frictions, which are included in the momentum equations, not only affect pressure drops along the fluid flow, but also have great effects for the numerical stability of the codes. In this study, the interfacial and wall frictions are reviewed for the commonly applied thermal-hydraulic system analysis codes, i.e. RELAP5-3D, MARS-3D, TRAC-M, and CATHARE.

  9. Comparison for the interfacial and wall friction models in thermal-hydraulic system analysis codes

    International Nuclear Information System (INIS)

    The average equations employed in the current thermal hydraulic analysis codes need to be closed with the appropriate models and correlations to specify the interphase phenomena along with fluid/structure interactions. This includes both thermal and mechanical interactions. Among the closure laws, an interfacial and wall frictions, which are included in the momentum equations, not only affect pressure drops along the fluid flow, but also have great effects for the numerical stability of the codes. In this study, the interfacial and wall frictions are reviewed for the commonly applied thermal-hydraulic system analysis codes, i.e. RELAP5-3D, MARS-3D, TRAC-M, and CATHARE

  10. Comparative risks of hydraulic, thermal and nuclear work in a large electrical utility

    International Nuclear Information System (INIS)

    Data is presented on the fatalities and accidents that have occurred in the construction and operation of hydraulic, thermal and nuclear generating facilities brought into service and operated by Ontario Hydro, a large Canadian utility, in the period 1970 to 1979. Results are also given of a prospective cohort epidemiological study of thermal and nuclear station workers which was begun in 1974. (author)

  11. Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10

    International Nuclear Information System (INIS)

    Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)

  12. Current and anticipated uses of the thermal hydraulics codes at the NRC

    Energy Technology Data Exchange (ETDEWEB)

    Caruso, R.

    1997-07-01

    The focus of Thermal-Hydraulic computer code usage in nuclear regulatory organizations has undergone a considerable shift since the codes were originally conceived. Less work is being done in the area of {open_quotes}Design Basis Accidents,{close_quotes}, and much more emphasis is being placed on analysis of operational events, probabalistic risk/safety assessment, and maintenance practices. All of these areas need support from Thermal-Hydraulic computer codes to model the behavior of plant fluid systems, and they all need the ability to perform large numbers of analyses quickly. It is therefore important for the T/H codes of the future to be able to support these needs, by providing robust, easy-to-use, tools that produce easy-to understand results for a wider community of nuclear professionals. These tools need to take advantage of the great advances that have occurred recently in computer software, by providing users with graphical user interfaces for both input and output. In addition, reduced costs of computer memory and other hardware have removed the need for excessively complex data structures and numerical schemes, which make the codes more difficult and expensive to modify, maintain, and debug, and which increase problem run-times. Future versions of the T/H codes should also be structured in a modular fashion, to allow for the easy incorporation of new correlations, models, or features, and to simplify maintenance and testing. Finally, it is important that future T/H code developers work closely with the code user community, to ensure that the code meet the needs of those users.

  13. Steady-state thermal-hydraulic of pebble bed blanket on hybrid reactor

    International Nuclear Information System (INIS)

    This paper gives thermal-hydraulic studies of pebble bed blanket on Hybrid Reactor. The concept of whole pebble bed blanket and the cooling methods are presented. The thermal-hydraulic characteristics of pebble bed are summarized. The theoretical model and code for solving heat transfer and flowing are presented. By using this code the calculation and analysis of thermal hydraulic of pebble bed Blanket of Hybrid Reactor are also given. In order to improve the flexibility, safety and economy, the authors select pebble beds not only to breed Tritium, but also to breed fission material and to multiply neutron. 5 MPa Helium is used as coolant and 0.05 MPa-0.1 MPa Helium is used as Purge gas. The heat transfer mechanisms of pebble bed are very complicated which include conduction, convection and radiation. In order to study the thermal-hydraulic of the bed, the authors just simply consider it as homogeneous and continuous binary phase medium as that used in the porous medium at the condition that the size of the bed is much greater than that of the balls. The coolant or the purge gas flowing through the bed is just considered existing a cooling source in the bed. It also significantly influences the effective conductivity's of the bed. Porous fraction, the main factor of the bed depends on the geometry position and parameters. From this model, one can obtain the thermal-hydraulic governing equations of the bed

  14. European liquid metal thermal-hydraulics R and D: present and future

    International Nuclear Information System (INIS)

    A large role is attributed in the future within the European Sustainable Nuclear Energy Technology Platform (SNE-TP) and especially the underlying European Sustainable Nuclear Industry Initiative (ESNII) to the application of fast reactors for sustainable nuclear energy production. Specifically, fast reactors are considered attractive because of their possibility to use natural resources efficiently and to reduce the volume and lifetime of nuclear waste. Currently four demonstration projects have a promising outlook in Europe, i.e. the ASTRID project in France, the MYRRHA project in Belgium, the ALFRED project developed in Europe and to be built in Romania, and the ELECTRA project in Sweden. Sodium and lead(-alloys) are envisaged as coolants for these reactors. Obviously, in the development of these reactors, thermal-hydraulics is recognized as a key challenge with emphasis on safety issues. This paper will discuss the present development status of liquid metal cooled reactor thermal-hydraulics as an outcome of the European 7. framework programme THINS (Thermal-Hydraulics for Innovative Nuclear Systems) project. The main project results with respect to liquid metal cooled reactors will be summarized, i.e. turbulence heat transfer model development, fuel assembly analysis, pool thermal-hydraulics, system behaviour, multi-phase physics, and multiscale thermal-hydraulics simulation. In conclusion, the main challenges for future developments will be indicated. Emphasis will be put on the important experimental and numerical challenges. (authors)

  15. Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Various codes are used for the thermal hydraulic analysis of nuclear reactors. The use of some codes among these is limited by user and some codes are not even open to general person. Thus, the use of alternative code is considered for some analysis. In this study, simple thermal hydraulic behaviors are analyzed using three codes to show that alternative codes are possible for the analysis of nuclear reactors. We established three models of the simple u-tube manometer using three different codes. RELAP5 (Reactor Excursion and Leak Analysis Program), SPACE (Safety and Performance Analysis CodE for nuclear power Plants), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are selected for this analysis. RELAP5 is widely used codes for the analysis of system behavior of PWRs. SPACE has been developed based on RELAP5 for the analysis of system behavior of PWRs and licensing of the code is in progress. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. The internal behavior of u-tube manometer was analyzed by RELAP5, SPACE and GOTHIC codes. The general transient behavior was similar among 3 codes. However, the stabilized status of the transient period analyzed by REPAP5 was different from the other codes. It would be resulted from the different physical models used in the other codes, which is specialized for the multi-phase thermal hydraulic behavior analysis

  16. Thermal-Hydraulic Analysis Tasks for ANAV NPPs in Support of Plant Operation and Control

    Directory of Open Access Journals (Sweden)

    L. Batet

    2007-11-01

    Full Text Available Thermal-hydraulic analysis tasks aimed at supporting plant operation and control of nuclear power plants are an important issue for the Asociación Nuclear Ascó-Vandellòs (ANAV. ANAV is the consortium that runs the Ascó power plants (2 units and the Vandellòs-II power plant. The reactors are Westinghouse-design, 3-loop PWRs with an approximate electrical power of 1000 MW. The Technical University of Catalonia (UPC thermal-hydraulic analysis team has jointly worked together with ANAV engineers at different levels in the analysis and improvement of these reactors. This article is an illustration of the usefulness of computational analysis for operational support. The contents presented were operational between 1985 and 2001 and subsequently changed slightly following various organizational adjustments. The paper has two different parts. In the first part, it describes the specific aspects of thermal-hydraulic analysis tasks related to operation and control and, in the second part, it briefly presents the results of three examples of analyses that were performed. All the presented examples are related to actual situations in which the scenarios were studied by analysts using thermal-hydraulic codes and prepared nodalizations. The paper also includes a qualitative evaluation of the benefits obtained by ANAV through thermal-hydraulic analyses aimed at supporting operation and plant control.

  17. Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yang Hoon; Jang, Mi Suk; Han, Kee Soo [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

    2015-05-15

    Various codes are used for the thermal hydraulic analysis of nuclear reactors. The use of some codes among these is limited by user and some codes are not even open to general person. Thus, the use of alternative code is considered for some analysis. In this study, simple thermal hydraulic behaviors are analyzed using three codes to show that alternative codes are possible for the analysis of nuclear reactors. We established three models of the simple u-tube manometer using three different codes. RELAP5 (Reactor Excursion and Leak Analysis Program), SPACE (Safety and Performance Analysis CodE for nuclear power Plants), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are selected for this analysis. RELAP5 is widely used codes for the analysis of system behavior of PWRs. SPACE has been developed based on RELAP5 for the analysis of system behavior of PWRs and licensing of the code is in progress. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. The internal behavior of u-tube manometer was analyzed by RELAP5, SPACE and GOTHIC codes. The general transient behavior was similar among 3 codes. However, the stabilized status of the transient period analyzed by REPAP5 was different from the other codes. It would be resulted from the different physical models used in the other codes, which is specialized for the multi-phase thermal hydraulic behavior analysis.

  18. Thermal hydraulic analyses of NPPs with VVER-440/213 for the PTS condition evaluation

    International Nuclear Information System (INIS)

    The paper presents the Nuclear Research Institute (NRI) Rez approach to selection of accident scenarios with risk of pressurized thermal shock (PTS) on reactor pressure vessel (RPV) and, therefore, loss of the RPV integrity. Some thermal hydraulic analyses of the phenomenon performed so far are described and main results are shown. Suitability of different types of thermal hydraulic computer codes to such analyses is discussed. Planned activities are also presented. Attention is focused to thermal hydraulics; problems of fracture mechanic analyses which naturally follow the thermal hydraulic analyses and give the final answer to the question of severity of given accident from the point of view of PTS are not treated in this paper. Identification of accident scenarios involving a risk of PTS is the first step of the PTS thermal-hydraulic and fracture mechanic analyses. The paper contains a discussion of such scenarios for NPPs with VVER-440/213 reactors. Attention is paid mainly to secondary circuit leakages. As an example, one accident scenario (main steam header rupture with multiple steamline break) is analysed by means of RELAP5 code, then the reactor downcomer and lower plenum are analysed by means of the CATHARE 2 vl.3 two-dimensional module, and, finally, velocity and temperature fields in one cold leg and a part of the reactor downcomer are calculated by means of 3-D thermal-hydraulic code FLUTAN 2 for 50 seconds after start of HPIS pumps and the case of primary coolant stagnation. Time dependent development of a cold plume in the calculational region is shown at various cross-sections. (author)

  19. Neutronics - thermal-hydraulics coupling: application to the helium-cooled fast reactor

    International Nuclear Information System (INIS)

    This thesis focuses on the study of interactions between neutron-kinetics and thermal-hydraulics. Neutron-kinetics allow to calculate the power in a nuclear reactor and the temperature evolution of materials where this power is deposited is known thanks to thermal-hydraulics. Moreover, when the temperatures evolve, the densities and cross sections change. These two disciplines are thus coupled. The first part of this work corresponds to the study and development of a method which allows to simulate transients in nuclear reactors and especially with a Monte-Carlo code for neutron-kinetics. An algorithm for the resolution of the neutron transport equation has been established and validated with a benchmark. In thermal-hydraulics, a porous media approach, based on another thesis, is considered. This gives the opportunity to solve the equations on the whole core without unconscionable computation time. Finally, a theoretical study has been performed on the statistical uncertainties which result from the use of a Monte-Carlo code and which spread from the reactivity to the power and from the power to the temperatures. The second part deals with the study of a misplaced control rod withdrawing in a GFR (helium-cooled fast reactor), a fourth generation reactor. Some models allowing to calculate neutron-kinetics and thermal-hydraulics in the core (which contains assemblies built up with fuel plates) were defined. In thermal-hydraulics, a model for the core based on the porous media approach and a fuel plate homogenization model have been set up. A similar homogenization model has been studied for neutron-kinetics. Finally, the control rod withdrawing transient where we can observe the power raising and the stabilisation by thermal feedback has been performed with the Monte-Carlo code Tripoli for neutron-kinetics and the code Trio-U for thermal-hydraulics. (author)

  20. Experimental verification of CFD and thermal hydraulics codes by quantitative flow visualisation

    International Nuclear Information System (INIS)

    Complex flow fields are encountered in many reactor components and processes. Measurement and analysis of various flow parameters are very important for optimal design, experimental determination of safety margins and verification of CFD and thermal hydraulics codes. Development of image capture hardware and digital image processing technique in Particle Image Velocimetry (PIV) has made possible to map complex flow fields instantaneously at thousands of points with very high temporal and spatial resolution. PIV is a non intrusive and very flexible technique. In this technique using synchronized operation of laser and CCD camera, seeded flow is illuminated by pulsing laser sheet and images of seeded particles are recorded on CCD camera. The displacement of the particles is measured in the plane of the image and used to determine the velocity of the flow. Image plane is divided into small interrogation regions. Velocity vectors are calculated with the help of cross correlated images obtained from two time exposures. This paper describes 2D PIV System used, flow mapping and verification of CFD codes for pipe flow, submerged jet, thermal stratification in water pool and Fluidic Flow Control Device (FFCD) proposed to be used in advanced accumulator of Emergency Core Cooling System (ECCS). (author)

  1. Finite mixture models for sensitivity analysis of thermal hydraulic codes for passive safety systems analysis

    International Nuclear Information System (INIS)

    Highlights: • Uncertainties of TH codes affect the system failure probability quantification. • We present Finite Mixture Models (FMMs) for sensitivity analysis of TH codes. • FMMs approximate the pdf of the output of a TH code with a limited number of simulations. • The approach is tested on a Passive Containment Cooling System of an AP1000 reactor. • The novel approach overcomes the results of a standard variance decomposition method. - Abstract: For safety analysis of Nuclear Power Plants (NPPs), Best Estimate (BE) Thermal Hydraulic (TH) codes are used to predict system response in normal and accidental conditions. The assessment of the uncertainties of TH codes is a critical issue for system failure probability quantification. In this paper, we consider passive safety systems of advanced NPPs and present a novel approach of Sensitivity Analysis (SA). The approach is based on Finite Mixture Models (FMMs) to approximate the probability density function (i.e., the uncertainty) of the output of the passive safety system TH code with a limited number of simulations. We propose a novel Sensitivity Analysis (SA) method for keeping the computational cost low: an Expectation Maximization (EM) algorithm is used to calculate the saliency of the TH code input variables for identifying those that most affect the system functional failure. The novel approach is compared with a standard variance decomposition method on a case study considering a Passive Containment Cooling System (PCCS) of an Advanced Pressurized reactor AP1000

  2. Finite mixture models for sensitivity analysis of thermal hydraulic codes for passive safety systems analysis

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, Francesco, E-mail: francesco.dimaio@polimi.it [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Nicola, Giancarlo [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Zio, Enrico [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Chair on System Science and Energetic Challenge Fondation EDF, Ecole Centrale Paris and Supelec, Paris (France); Yu, Yu [School of Nuclear Science and Engineering, North China Electric Power University, 102206 Beijing (China)

    2015-08-15

    Highlights: • Uncertainties of TH codes affect the system failure probability quantification. • We present Finite Mixture Models (FMMs) for sensitivity analysis of TH codes. • FMMs approximate the pdf of the output of a TH code with a limited number of simulations. • The approach is tested on a Passive Containment Cooling System of an AP1000 reactor. • The novel approach overcomes the results of a standard variance decomposition method. - Abstract: For safety analysis of Nuclear Power Plants (NPPs), Best Estimate (BE) Thermal Hydraulic (TH) codes are used to predict system response in normal and accidental conditions. The assessment of the uncertainties of TH codes is a critical issue for system failure probability quantification. In this paper, we consider passive safety systems of advanced NPPs and present a novel approach of Sensitivity Analysis (SA). The approach is based on Finite Mixture Models (FMMs) to approximate the probability density function (i.e., the uncertainty) of the output of the passive safety system TH code with a limited number of simulations. We propose a novel Sensitivity Analysis (SA) method for keeping the computational cost low: an Expectation Maximization (EM) algorithm is used to calculate the saliency of the TH code input variables for identifying those that most affect the system functional failure. The novel approach is compared with a standard variance decomposition method on a case study considering a Passive Containment Cooling System (PCCS) of an Advanced Pressurized reactor AP1000.

  3. Advanced Hydraulic Fracturing Technology for Unconventional Tight Gas Reservoirs

    Energy Technology Data Exchange (ETDEWEB)

    Stephen Holditch; A. Daniel Hill; D. Zhu

    2007-06-19

    The objectives of this project are to develop and test new techniques for creating extensive, conductive hydraulic fractures in unconventional tight gas reservoirs by statistically assessing the productivity achieved in hundreds of field treatments with a variety of current fracturing practices ranging from 'water fracs' to conventional gel fracture treatments; by laboratory measurements of the conductivity created with high rate proppant fracturing using an entirely new conductivity test - the 'dynamic fracture conductivity test'; and by developing design models to implement the optimal fracture treatments determined from the field assessment and the laboratory measurements. One of the tasks of this project is to create an 'advisor' or expert system for completion, production and stimulation of tight gas reservoirs. A central part of this study is an extensive survey of the productivity of hundreds of tight gas wells that have been hydraulically fractured. We have been doing an extensive literature search of the SPE eLibrary, DOE, Gas Technology Institute (GTI), Bureau of Economic Geology and IHS Energy, for publicly available technical reports about procedures of drilling, completion and production of the tight gas wells. We have downloaded numerous papers and read and summarized the information to build a database that will contain field treatment data, organized by geographic location, and hydraulic fracture treatment design data, organized by the treatment type. We have conducted experimental study on 'dynamic fracture conductivity' created when proppant slurries are pumped into hydraulic fractures in tight gas sands. Unlike conventional fracture conductivity tests in which proppant is loaded into the fracture artificially; we pump proppant/frac fluid slurries into a fracture cell, dynamically placing the proppant just as it occurs in the field. From such tests, we expect to gain new insights into some of the critical

  4. 75 FR 69140 - NUREG-1953, Confirmatory Thermal-Hydraulic Analysis To Support Specific Success Criteria in the...

    Science.gov (United States)

    2010-11-10

    ... COMMISSION NUREG-1953, Confirmatory Thermal-Hydraulic Analysis To Support Specific Success Criteria in the...- Hydraulic Analysis to Support Specific Success Criteria in the Standardized Plant Analysis Risk Models...-Hydraulic Analysis to Support Specific Success Criteria in the Standardized Plant Analysis Risk...

  5. Development and study of thermal-hydraulic code for spiral-space rods assembly

    International Nuclear Information System (INIS)

    Spiral-spacer fuel assembly usually adopts helical fins or wire wrap fuel elements. Compare with the tradition- al PWR fuel rods, spiral spacers make the thermal hydraulic phenomena in sub-channels very complicated. The paper preliminary studied the influence of the spiral spacer to the thermal-hydraulic performance, there is no suitable code to study these affect. A new code named CANAL/CMS was developed base on the VVER code. Using the new code, investigation has been carried out for the influence of the helical fins. Systemic study shows that the impact of the helical fins to the thermal hydraulic of the bundle is great; they improve the ability of the heat transfer of the fuel elements to a certain extent, and the pressure drop add little; the long helical spacer will reduce the pressure drop, but it is bad for CHF. (authors)

  6. Thermal hydraulic analysis for the Oregon State TRIGA reactor using RELAP5-3D

    International Nuclear Information System (INIS)

    Thermal hydraulic analyses have being conducted at Oregon State University (OSU) in support of the conversion of the OSU TRIGA reactor (OSTR) core from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel as part of the Reduced Enrichment for Research and Test Reactors program. The goals of the thermal hydraulic analyses were to calculate natural circulation flow rates, coolant temperatures and fuel temperatures as a function of core power for both the HEU and LEU cores; calculate peak values of fuel temperature, cladding temperature, surface heat flux as well as departure from nuclear boiling ratio (DNBR) for steady state and pulse operation; and perform accident analyses for the accident scenarios identified in the OSTR safety analysis report. RELAP5-3D Version 2.4.2 was implemented to develop a model for the thermal hydraulic study. The OSTR core conversion is planned to take place in late 2008. (author)

  7. Thermal hydraulic calculations of the IRT-200 reactor with LEU IRT-4M, Sofia

    International Nuclear Information System (INIS)

    Neutronic calculations of the IRT-200 research reactor with fuel assemblies (FA) of the IRT-4M type, containing low enriched uranium (19.75 %), performed by the Institute for Nuclear Research and Nuclear Energy (INRNE) jointly with the RERTR Program at Argonne National Laboratory (ANL), confirmed justness of selection of its initial core configuration. On the base of neutronic calculation results thermal hydraulic calculations were done by PLTEMP code. Three possible operational regimes have been considered and for each one of them the margin coefficient of the water onset of nucleate boiling (ONB) on the fuel element surface for the maximum power density fuel assembly has been determined. The calculations have been carried out for the core water inlet temperature of 45 deg. C The thermal hydraulic calculations demonstrated satisfaction of thermal hydraulic safety margins requirements even at 1 MW power level. (author)

  8. Development of Computer Program for Whole Core Thermal-Hydraulic Analysis of Fast Reactors

    International Nuclear Information System (INIS)

    A whole core thermal-hydraulic analysis program ACT was developed for the purpose of evaluating detailed in-core thermal-hydraulic phenomena of sodium cooled fast reactors under various reactor operation conditions. ACT consists of four kinds of calculation modules, i.e., fuel-assembly, inter-wrapper gap (core barrel), upper plenum and heat transport system modules. The latter two modules give proper boundary conditions for the reactor core thermal-hydraulic analysis. These four modules are coupled with each other by using MPI and calculate simultaneously on a cluster workstation. ACT was applied to analyzing a sodium experiment performed at JNC, which simulated the natural circulation decay heat removal under PRACS and DRACS operation condition. In the experiment, not only inter-wrapper flows but also reverses flows in the fuel assemblies were observed. ACT succeeded in simulating such complicated phenomena. (authors)

  9. Benchmark of SIMULATE5 thermal hydraulics against the Frigg and NUPEC full bundle test experiments

    International Nuclear Information System (INIS)

    SIMULATE5 is Studsvik Scandpower's next generation nodal code. The core portion of the thermal hydraulic models of PWR and BWRs are treated as essentially identical, with each assembly having an active channel and a number of parallel water channels. In addition, the BWR assembly may be divided into four radial sub-assemblies. For natural circulation reactors, the BWR thermal hydraulic model is capable of modeling an entire vessel loop: core, chimney, upper plenum, standpipes, steam separators, downcomer, recirculation pumps, and lower plenum. This paper presents results of the validation of the BWR thermal hydraulic model against: (1) pressure drop data measured in the Frigg and NUPEC test facilities; (2) void fraction distribution measured in the Frigg and NUPEC loops; (3) quarter-assembly void fraction measured in the NUPEC experiments and (4) natural and forced circulation flow measurements in the Frigg loop. (author)

  10. IMPULSE - advanced nuclear thermal propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Ivanenok, J.F. III; Wett, J.F. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1993-12-31

    The IMPULSE nuclear thermal rocket concept provides an evolutionary step toward high thrust-to-weight and specific impulse over a wide operating range. Most of the components and features of the concept are based on demonstrated or proven technology from the NER VA/Rover program. The performance increase is due to the use of a new solid nuclear fuel shape. The new fuel shape provides a large flow area while maintaining flow control and eliminating hot spots due to fuel-to-fuel contact. The control and eliminating hot spots due to fuel-to-fuel contact. The IMPULSE reactor utilizes a multi-pass, series flow configuration to provide excess turbine power while improving the thermal efficiency of the overall system. This configuration also provides a large area for moderator. The IMPULSE concept can provide a specific impulse of up to 1000 seconds and trust to weight ratios approaching 40. The improved performance will reduce the Initial Mass In Low Earth Orbit (IMLEO) and provide a consequent reduction in launch costs and logistics problems.

  11. Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

    International Nuclear Information System (INIS)

    Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has been restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory's (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed a Readiness Assessment in November 2009. The HSIS is a key component of the ATR National Scientific User Facility (NSUF) operated by Battelle Energy Alliance, LLC and is available to a wide variety of university researchers for nuclear fuels and materials experiments as well as medical isotope research and production.

  12. Development of water packing mitigation scheme for MARS 3- dimensional thermal-hydraulic module

    International Nuclear Information System (INIS)

    Water packing mitigation scheme was developed to enhance the numerical stability and calculational efficiency of MARS 3-dimensional thermal-hydraulic module. The water packing phenomena is unphysical pressure spike which occurs in a two-phase system thermal-hydraulic code using Eulerian finite difference method. Great velocities developed from large pressure spike slow down the calculation efficiency due to the stability limit. Also, large pressure spike and subsequent low pressure can make errors in thermodynamic state table search. The developed water packing mitigation scheme was implemented in MARS3D module. It is shown from the results of some benchmark problema that numerical stability and calculational efficiency were improved

  13. SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries

  14. Proceedings: MAAP thermal-hydraulic qualifications and guidelines for plant application workshop

    International Nuclear Information System (INIS)

    This workshop was held in Houston, November 15--16, 1990. It was organized as a technology-transfer activity, intended to provide utilities and others in the industry with the results of much-needed plan applications. The workshop focused on the MAAP thermal-hydraulic qualification program, selection of risk-dominant scenarios and thermal-hydraulic phenomena, PWR and BWR assessments, plant-specific applications, quantification of success criteria and probabilistic risk assessment (PRA) applications, and use of results for IPEs and application guidelines. Individual papers have been cataloged separately

  15. Neutronic and thermal hydraulic analysis for production of fission molybdenum-99 at Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    Neutronic and thermal hydraulic analysis for the fission molybdenum-99 production at PARR-1 has been performed. Low enriched uranium foil (235U) will be used as target material. Annular target designed by ANL (USA) will be irradiated in PARR-1 for the production of 100 Ci of molybdenum-99 at the end of irradiation, which will be sufficient to prepare required 99Mo/99mTc generators at PINSTECH and its supply in the country. Neutronic and thermal hydraulic analysis were performed using various codes. Data shows that annular targets can be safely irradiated in PARR-1 for production of required amount of fission molybdenum-99

  16. Thermal - hydraulic analysis of pressurizer water reactors using the model of open lateral boundary

    International Nuclear Information System (INIS)

    A computational method is developed for thermal-hydraulic analysis, where the channel may be analysed by more than one independent steps of calculation. This is made possible by the incorporation of the model of open lateral boundary in the code COBRA-IIIP, which permits the determination of the subchannel of an open lattice PWR core in a multi-step calculation. The thermal-hydraulic code COBRA-IIIP, developed at the Massachusetts Institute of Technology, is used as the basic model for this study. (Author)

  17. Methods and programs of thermal hydraulic calculations of fast reactor fuel assemblies

    International Nuclear Information System (INIS)

    The methods and computer codes for calculating the velocity and temperature distributions in fast reactor fuel assemblies are described and analyzed. Three levels of thermal hydraulic analysis of fuel element bundles can be distinguished, viz.: analysis of local characteristics (finite element method, finite difference method), subchannel analysis (lumped parameter method), and analysis of characteristics averaged over volumes (porous body model). The possibilities of the existing computer codes and methods are demonstrated and conclusions regarding the future development of methods of and codes for thermal hydraulic analysis of fuel assemblies are presented. (author). 102 figs., 17 tabs., 256 refs

  18. Current and anticipated uses of thermal hydraulic codes at the Japan Atomic Energy Research Institute

    Energy Technology Data Exchange (ETDEWEB)

    Akimoto, Hajime; Kukita; Ohnuki, Akira [Japan Atomic Energy Research Institute, Ibaraki (Japan)

    1997-07-01

    The Japan Atomic Energy Research Institute (JAERI) is conducting several research programs related to thermal-hydraulic and neutronic behavior of light water reactors (LWRs). These include LWR safety research projects, which are conducted in accordance with the Nuclear Safety Commission`s research plan, and reactor engineering projects for the development of innovative reactor designs or core/fuel designs. Thermal-hydraulic and neutronic codes are used for various purposes including experimental analysis, nuclear power plant (NPP) safety analysis, and design assessment.

  19. Thermal-hydraulics and safety concepts of supercritical water cooled reactors

    International Nuclear Information System (INIS)

    The paper summarizes the status of safety system development for supercritical water cooled reactors and of thermal-hydraulic codes needed to analyze them. While active safety systems are well understood today and expected to perform as required, the development of passive safety systems will still need further optimization. Depressurization transients have successfully been simulated with some codes by a pseudo-two-phase flow simulation of supercritical water. Open issues of thermal-hydraulic codes include modeling of deteriorated heat transfer in one-dimensional system codes and predictions of heat transfer during depressurization transients from supercritical to sub-critical conditions. (author)

  20. Thermal hydraulic analysis of the multipurpose research reactor RMB using a RELAP5 model

    International Nuclear Information System (INIS)

    The Multipurpose Brazilian Reactor (RMB) will be an open pool multipurpose research reactor using low enriched uranium fuel (LEU). This paper presents the RMB nodalization and the first thermal hydraulic results of steady state calculations using the RELAP5-MOD3.3 code. Several current investigations have shown that RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research systems with good predictions in spite of such code was initially projected to studies of commercial nuclear power plants. (author)

  1. Light-water-reactor coupled neutronic and thermal-hydraulic codes

    International Nuclear Information System (INIS)

    An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three steady state codes (PDQ7, NODE-P/B and SIMULATE) and four dynamic codes (BNL-TWIGL, MEKIN, RAMONA-3B, RETRAN-02). Speculation as to future trends with such codes is also presented

  2. Thermal hydraulic behavior during main steam line break in PTS evaluation of Kori Unit 1

    International Nuclear Information System (INIS)

    The present study primarily aims to evaluate the system thermal hydraulic and thermal mixing behavior in downcomer of a postulated Main Steam Line Break (MSLB) as one of the major PTS (Pressurized Thermal Shock) initiators in PTS Evaluation of Kori Unit 1. For this purpose, the MSLB event sequences were reviewed and the most severe event sequence was selected based on the peer review on design and operational features. Based on the present design and operation condition of Kori Unit 1, a base calculation of the most severe sequence of MSLB events is conducted using RELAP5 and the sensitivity of the thermal-hydraulic mixing in downcomer was also investigated. The current result shows an overall downcomer fluid cooling from 558degK to 436degK. From the sensitivity analysis results, it is found that the thermal mixing could be affected by the downcomer modeling. (author)

  3. Development of numerical procedure for thermal hydraulic design of nuclear reactors with advanced two-fluid model (2). Applicability of turbulent dispersion force model for middle diameter vertical pipe

    International Nuclear Information System (INIS)

    Advanced two-fluid model has been developed in Japan Atomic Energy Agency (JAEA). In the model, an interface tracking method is combined with the two-fluid model. Liquid clusters and bubbles larger than a computational cell are calculated using the interface tracking method, and those smaller than a cell are simulated by the two-fluid model. The turbulent dispersion force term is one of the most important constitutive equations for advanced two-fluid model. In the past, we have developed the new model for turbulent dispersion force and verified the developed model using the 3-dimentional two-fluid model code ACE-3D developing by JAEA, and the comparisons between the results of the analyses and air-water two-phase flow experiments in 200 mm diameter pipe. In this study, to verify applicability of middle diameter vertical pipe using the developed model, we compared between the analyses of results and air-water two-phase flow experiments in 38 mm diameter pipe. (author)

  4. Deterministic and Monte Carlo transport models with thermal-hydraulic feedback

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

    This paper gives an overview of recent developments concerning deterministic transport and Monte Carlo methods with thermal-hydraulic feedback. The timedependent 3D discrete ordinates transport code TORT-TD allows pin-by-pin analyses of transients using few energy groups and anisotropic scattering by solving the timedependent transport equation using the unconditionally stable implicit method. To account for thermal-hydraulic feedback, TORT-TD has been coupled with the system code ATHLET. Applications to, e.g., a control rod ejection in a 2 x 2 PWR fuel assembly arrangement demonstrate the applicability of the coupled code TORT-TD/ATHLET for test cases. For Monte Carlo steady-state calculations with nuclear point data and thermalhydraulic feedback, MCNP has been prepared to incorporate thermal-hydraulic parameters. As test case has been chosen the uncontrolled steady state of the 2 x 2 PWR fuel assembly arrangement for which the thermal-hydraulic parameter distribution has been obtained from a preceding coupled TORT-TD/ATHLET analysis. The result demonstrates the applicability of MCNP to problems with spatial distributions of thermal-fluiddynamic parameters. The comparison with MCNP results confirms that the accuracy of deterministic transport calculations with pin-wise homogenised few-group cross sections is comparable to Monte Carlo simulations. The presented cases are considered as a pre-stage of performing calculations of larger configurations like a quarter core which is in preparation. (orig.)

  5. Deterministic and Monte Carlo transport models with thermal-hydraulic feedback

    International Nuclear Information System (INIS)

    This paper gives an overview of recent developments concerning deterministic transport and Monte Carlo methods with thermal-hydraulic feedback. The timedependent 3D discrete ordinates transport code TORT-TD allows pin-by-pin analyses of transients using few energy groups and anisotropic scattering by solving the timedependent transport equation using the unconditionally stable implicit method. To account for thermal-hydraulic feedback, TORT-TD has been coupled with the system code ATHLET. Applications to, e.g., a control rod ejection in a 2 x 2 PWR fuel assembly arrangement demonstrate the applicability of the coupled code TORT-TD/ATHLET for test cases. For Monte Carlo steady-state calculations with nuclear point data and thermalhydraulic feedback, MCNP has been prepared to incorporate thermal-hydraulic parameters. As test case has been chosen the uncontrolled steady state of the 2 x 2 PWR fuel assembly arrangement for which the thermal-hydraulic parameter distribution has been obtained from a preceding coupled TORT-TD/ATHLET analysis. The result demonstrates the applicability of MCNP to problems with spatial distributions of thermal-fluiddynamic parameters. The comparison with MCNP results confirms that the accuracy of deterministic transport calculations with pin-wise homogenised few-group cross sections is comparable to Monte Carlo simulations. The presented cases are considered as a pre-stage of performing calculations of larger configurations like a quarter core which is in preparation. (orig.)

  6. Evaluation of thermal-hydraulic parameter uncertainties in a TRIGA research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Costa, Antonio C.L.; Ladeira, Luiz C.D.; Rezende, Hugo C., E-mail: amir@cdtn.br, E-mail: aclc@cdtn.br, E-mail: lcdl@cdtn.br, E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    Experimental studies had been performed in the TRIGA Research Nuclear Reactor of CDTN/CNEN to find out the its thermal hydraulic parameters. Fuel to coolant heat transfer patterns must be evaluated as function of the reactor power in order to assess the thermal hydraulic performance of the core. The heat generated by nuclear fission in the reactor core is transferred from fuel elements to the cooling system through the fuel-cladding (gap) and the cladding to coolant interfaces. As the reactor core power increases the heat transfer regime from the fuel cladding to the coolant changes from single-phase natural convection to subcooled nucleate boiling. This paper presents the uncertainty analysis in the results of the thermal hydraulics experiments performed. The methodology used to evaluate the propagation of uncertainty in the results was done based on the pioneering article of Kline and McClintock, with the propagation of uncertainties based on the specification of uncertainties in various primary measurements. The uncertainty analysis on thermal hydraulics parameters of the CDTN TRIGA fuel element is determined, basically, by the uncertainty of the reactor's thermal power. (author)

  7. A domain decomposition methodology through alternate dissections for coupled neutronic and thermal-hydraulic analyses in Cobaya3

    International Nuclear Information System (INIS)

    Nowadays, coupled 3D neutron-kinetics and thermal-hydraulic core calculations are performed by applying a radial average channel approach using a meshing of one quarter of assembly in the best case. This approach does not take into account the sub-channels effects due to the averaging of the physical fields and the loose of heterogeneity in the thermal-hydraulic modelization. Therefore the models do not have enough resolution to predict those sub-channels effects which are important for the fuel design safety margins, because it is in the local scale, where we can search the hottest pellet or the maximum heat flux. The UPM advanced multi-scale neutron-kinetics and thermal-hydraulics methodologies being implemented in COBAYA3 include domain decomposition by alternate core dissections for the local 3D fine-mesh scale problems (pin cells/sub-channels) and an analytical nodal diffusion solver for the coarse mesh scale coupled with the thermal-hydraulic using a modelization of one channel per assembly or per quarter of assembly. In this work, we address the domain decomposition by the alternate core dissections methodology applied to solve coupled 3D neutronic/thermal-hydraulic (N-TH) problems at the fine-mesh scale. The N-TH coupling at the cell-sub-channel scale allows the treatment of the effects of the detailed TH feedbacks on cross-sections, thus resulting in better estimates of the local safety margins at the pin level. If we try to perform a sub-channel-by-sub-channel coupled N-TH calculation for a full PWR core in order to get all the detailed effects, the need of memory and CPU calculation time will be huge and even in the case we achieve to converge an steady state scenario at nominal conditions, which is a non-trivial case, the main problem come to us when we introduce some perturbations in the system (i.e. insertion of control rods) that makes the problem asymmetric and by far more heterogeneous. It would be very difficult to follow the propagation of the

  8. A domain decomposition methodology through alternate dissections for coupled neutronic and thermal-hydraulic analyses in Cobaya3

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, Javier; Aragones, Jose Maria; Cuervo, Diana [Nuclear Engineering Department, E.T.S.I. Industriales, Universidad Politecnica de Madrid, c/ Jose Gutierrez Abascal, 2, 28006 Madrid (Spain)

    2010-07-01

    Nowadays, coupled 3D neutron-kinetics and thermal-hydraulic core calculations are performed by applying a radial average channel approach using a meshing of one quarter of assembly in the best case. This approach does not take into account the sub-channels effects due to the averaging of the physical fields and the loose of heterogeneity in the thermal-hydraulic modelization. Therefore the models do not have enough resolution to predict those sub-channels effects which are important for the fuel design safety margins, because it is in the local scale, where we can search the hottest pellet or the maximum heat flux. The UPM advanced multi-scale neutron-kinetics and thermal-hydraulics methodologies being implemented in COBAYA3 include domain decomposition by alternate core dissections for the local 3D fine-mesh scale problems (pin cells/sub-channels) and an analytical nodal diffusion solver for the coarse mesh scale coupled with the thermal-hydraulic using a modelization of one channel per assembly or per quarter of assembly. In this work, we address the domain decomposition by the alternate core dissections methodology applied to solve coupled 3D neutronic/thermal-hydraulic (N-TH) problems at the fine-mesh scale. The N-TH coupling at the cell-sub-channel scale allows the treatment of the effects of the detailed TH feedbacks on cross-sections, thus resulting in better estimates of the local safety margins at the pin level. If we try to perform a sub-channel-by-sub-channel coupled N-TH calculation for a full PWR core in order to get all the detailed effects, the need of memory and CPU calculation time will be huge and even in the case we achieve to converge an steady state scenario at nominal conditions, which is a non-trivial case, the main problem come to us when we introduce some perturbations in the system (i.e. insertion of control rods) that makes the problem asymmetric and by far more heterogeneous. It would be very difficult to follow the propagation of the

  9. Quasi-one-dimensional modeling of subcooled boiling thermal-hydraulic characteristics

    International Nuclear Information System (INIS)

    Full text of publication follows: The study of thermal-hydraulic characteristics in two phase non-homogeneous flows by means of one dimensional numeric mathematical models remains at present one of the most popular approach applied in calculations practice for contemporary high-temperature energy equipment in traditional and nuclear power engineering. Such study becomes more actual with the creating of advanced models describing non-homogeneous thermal-hydraulic processes for 'best estimated' codes. Physical-mathematical formulations are, as a rule, classified by the empiric contents and accepted assumption levels of the subcooled boiling models. The less restricted of them are the mechanistic models based on the use of vapor generation and condensation functions. The accounting for non-homogeneous spatial effects is being fulfilled, at that, by selecting 'fitting' constants in the above mentioned functions. By 'fitting' constants, usually, are implied constants introduced by authors, selected under the condition of optimal description of the processed experimental data by a one-dimensional calculation model. The limited working diapason of such applied methods, as well as possible combinations of regime parameters substantially influencing on discrepancies between experimental and calculated data, authors address to the display of complex spatial (two-dimensional in axisymmetric case) effects in channel without, as a rule, giving corresponding quantitative models. A correct way, allowing to take into account the real two-dimensional flow character in one-dimensional models, is in employing the technique of spatial averaging for local non-homogeneities, resulting in additional distribution parameters (DP) and form factors (FF) in one-dimensional conservation laws for two-phase flow. By its physical interpretation, DP and FF represent a measure of the real profile deviation from the flat case and represent correction coefficient to the left and right parts of one

  10. Development of a transient thermal-hydraulic code for analysis of China Demonstration Fast Reactor

    International Nuclear Information System (INIS)

    Highlights: ► A transient thermal-hydraulic code is developed for analysis of CDFR. ► The code to code validation shows good accuracy and reliability of the code. ► Multiple-channel model is applied to depict the core. ► Compressible homogenous flow model is used for the two-phase flow of sodium. - Abstract: The transient thermal-hydraulic code THACOS is under development for analysis of China Demonstration Fast Reactor. Applying modular technology, the code contains the core module, the pump module, the sodium pool module and the heat exchanger module and each module could operate separately. It can provide one-dimensional thermal-hydraulic simulation for the primary sodium coolant loop. The point reactor kinetics equations with six-group delayed neutrons have been applied to calculate the core power considering reactivity feedbacks caused by the Doppler effect, coolant density, axial expansion of fuel rods and radial expansion of the core. Multiple-channel model is applied to depict the core. Compressible homogenous flow model is used for the two-phase flow of sodium. The calculated results show that sodium boiling will occur quickly under the ULOF accident without any shutdown rods insertion. While, with the insertion of three hydraulically suspended shutdown rods, the core could be shut down safely and boiling will not occur in a short period of time. Obviously, the passive hydraulically suspended shutdown rods could keep the core safe under ULOF accidents

  11. Thermal hydraulic analysis of main-steam-line-break accidents as potential initiators for reactor vessel pressurized thermal shock

    International Nuclear Information System (INIS)

    Results are presented from two thermal hydraulic analysis of postulated main-steam-line breaks for the Oconee nuclear power plant. One calculation assumes runaway feedwater supply, whereas normal feedwater management is used in the other. The analyses were performed with the TRAC-PD2 code. The objective was to provide primary coolant temperature and pressure histories to assist in evaluating possible reactor-vessel pressurized thermal-shock concerns

  12. Uncertainty Evaluation of the SFR Subchannel Thermal-Hydraulic Modeling Using a Hot Channel Factors Analysis

    International Nuclear Information System (INIS)

    In an SFR core analysis, a hot channel factors (HCF) method is most commonly used to evaluate uncertainty. It was employed to the early design such as the CRBRP and IFR. In other ways, the improved thermal design procedure (ITDP) is able to calculate the overall uncertainty based on the Root Sum Square technique and sensitivity analyses of each design parameters. The Monte Carlo method (MCM) is also employed to estimate the uncertainties. In this method, all the input uncertainties are randomly sampled according to their probability density functions and the resulting distribution for the output quantity is analyzed. Since an uncertainty analysis is basically calculated from the temperature distribution in a subassembly, the core thermal-hydraulic modeling greatly affects the resulting uncertainty. At KAERI, the SLTHEN and MATRA-LMR codes have been utilized to analyze the SFR core thermal-hydraulics. The SLTHEN (steady-state LMR core thermal hydraulics analysis code based on the ENERGY model) code is a modified version of the SUPERENERGY2 code, which conducts a multi-assembly, steady state calculation based on a simplified ENERGY model. The detailed subchannel analysis code MATRA-LMR (Multichannel Analyzer for Steady-State and Transients in Rod Arrays for Liquid Metal Reactors), an LMR version of MATRA, was also developed specifically for the SFR core thermal-hydraulic analysis. This paper describes comparative studies for core thermal-hydraulic models. The subchannel analysis and a hot channel factors based uncertainty evaluation system is established to estimate the core thermofluidic uncertainties using the MATRA-LMR code and the results are compared to those of the SLTHEN code

  13. Uncertainty Evaluation of the SFR Subchannel Thermal-Hydraulic Modeling Using a Hot Channel Factors Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Sun Rock; Cho, Chung Ho; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    In an SFR core analysis, a hot channel factors (HCF) method is most commonly used to evaluate uncertainty. It was employed to the early design such as the CRBRP and IFR. In other ways, the improved thermal design procedure (ITDP) is able to calculate the overall uncertainty based on the Root Sum Square technique and sensitivity analyses of each design parameters. The Monte Carlo method (MCM) is also employed to estimate the uncertainties. In this method, all the input uncertainties are randomly sampled according to their probability density functions and the resulting distribution for the output quantity is analyzed. Since an uncertainty analysis is basically calculated from the temperature distribution in a subassembly, the core thermal-hydraulic modeling greatly affects the resulting uncertainty. At KAERI, the SLTHEN and MATRA-LMR codes have been utilized to analyze the SFR core thermal-hydraulics. The SLTHEN (steady-state LMR core thermal hydraulics analysis code based on the ENERGY model) code is a modified version of the SUPERENERGY2 code, which conducts a multi-assembly, steady state calculation based on a simplified ENERGY model. The detailed subchannel analysis code MATRA-LMR (Multichannel Analyzer for Steady-State and Transients in Rod Arrays for Liquid Metal Reactors), an LMR version of MATRA, was also developed specifically for the SFR core thermal-hydraulic analysis. This paper describes comparative studies for core thermal-hydraulic models. The subchannel analysis and a hot channel factors based uncertainty evaluation system is established to estimate the core thermofluidic uncertainties using the MATRA-LMR code and the results are compared to those of the SLTHEN code

  14. Thermal-hydraulic analysis of SMART steam generator tube rupture using TASS/SMR-S code

    International Nuclear Information System (INIS)

    Highlights: ► The analysis was performed from the viewpoint of primary coolant leakage. ► The thermal hydraulic responses and the maximum leakage have been identified. ► There is no direct release into the atmosphere caused by an SGTR accident. ► SMART safety system works well against an SGTR accident. - Abstract: A steam generator tube rupture (SGTR) accident analysis for SMART was performed using the TASS/SMR-S code. SMART with a rated thermal power of 330 MWt has been developed at the Korea Atomic Energy Research Institute. The TASS/SMR-S code can analyze the thermal hydraulic phenomena of SMART in a full range of reactor operating conditions. An SGTR is one of the most important accidents from a thermal hydraulic and radiological viewpoint. A conservative analysis against a SMART SGTR was performed. The major concern of this analysis is to find the thermal hydraulic responses and maximum leakage amount from a primary to a secondary side caused by an SGTR accident. A sensitivity study searching for the conservative thermal hydraulic conditions, break locations, reactivity and other conditions was performed. The dominant parameters related with the integral leak are the high RCS pressure, low core inlet coolant temperature and low break location of the SG cassette. The largest integral leak comes to 28 tons in the most conservative case during 1 h. But there is no direct release into the atmosphere because the secondary system pressure is maintained with a sufficient margin for the design pressure. All leaks go to the condenser. The analysis results show that the primary and secondary system pressures are maintained below the design pressure and the SMART safety system is working well against an SGTR accident

  15. Thermal-hydraulic calculation and test for irradiation target assembly

    International Nuclear Information System (INIS)

    In the paper, the geometry and mathematics model of the irradiation Target Assembly are built. COBRAIII C/MIT cod is used to calculate the flow rate and pressure drop curve of the irradiation Target Assembly, and the error of theoretic result and hydraulic test is 1.32%. The maximum temperatures on the surface and core of Irradiation Target Assembly are 66.6 degree C and 72.7 degree C, which are less than the design limited value. The MDNBR (minimum departure from nucleate boiling ratio) is also calculated in the paper, which is 5.72 and more than the design limited value. The calculation result has significance for the safety analysis of Irradiation Target Assembly. (authors)

  16. JPL Advanced Thermal Control Technology Roadmap - 2008

    Science.gov (United States)

    Birur, Gaj

    2008-01-01

    This slide presentation reviews the status of thermal control technology at JPL and NASA.It shows the active spacecraft that are in vairous positions in the solar syatem, and beyond the solar system and the future missions that are under development. It then describes the challenges that the past missions posed with the thermal control systems. The various solutions that were implemented duirng the decades prior to 1990 are outlined. A review of hte thermal challenges of the future misions is also included. The exploration plan for Mars is then reviewed. The thermal challenges of the Mars Rovers are then outlined. Also the challenges of systems that would be able to be used in to explore Venus, and Titan are described. The future space telescope missions will also need thermal control technological advances. Included is a review of the thermal requirements for manned missions to the Moon. Both Active and passive technologies that have been used and will be used are reviewed. Those that are described are Mechanically Pumped Fluid Loops (MPFL), Loop Heat Pipes, an M3 Passive Cooler, Heat Siwtch for Space and Mars surface applications, phase change material (PCM) technology, a Gas Gap Actuateor using ZrNiH(x), the Planck Sorption Cooler (PCS), vapor compression -- Hybrid two phase loops, advanced pumps for two phase cooling loops, and heat pumps that are lightweight and energy efficient.

  17. Thermal-hydraulic characteristics of the RBMK control and protection system channels

    International Nuclear Information System (INIS)

    The thermal-hydraulic characteristics of the RBMK-1000 control and protection system channel with rod cluster control have been calculated under different operational disturbance regimes. It has been shown that the temperature of the rod cluster control structural materials increases considerably if loss of coolant occurs. The critical element is the sleeve made of CAB1 aluminum alloy

  18. Analysis of the flow dynamics in the multiloop thermal-hydraulic system

    International Nuclear Information System (INIS)

    The method for identification of the closed loops in the thermal-hydraulics system is described. The mathematical models, previously developed for the single loop system, have been extended to treat multiloop pipelines. The configuration of the network is practically deliberate. The model has been tested by the simulation of the asymmetric transient in the two loop PWR. (author)

  19. Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

    International Nuclear Information System (INIS)

    Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor

  20. Survey of thermal-hydraulic models of commercial nuclear power plants

    International Nuclear Information System (INIS)

    A survey of the thermal-hydraulic models of nuclear power plants has been performed to identify the NRC's current analytical capabilities for critical event response. The survey also supports ongoing research for accident management. The results of the survey are presented here. The PC database which records detailed data on each model is described

  1. Thermal-hydraulic assessment of concrete storage cubicle with horizontal 3013 canisters

    International Nuclear Information System (INIS)

    The FIDAP computer code was used to perform a series of analyses to assess the thermal-hydraulic performance characteristics of the concrete plutonium storage cubicles, as modified for the horizontal placement of 3013 canisters. Four separate models were developed ranging from a full height model of the storage cubicle to a very detailed standalone model of a horizontal 3013 canister

  2. Coupled neutronics and thermal hydraulics analysis for PWR with MCNP5 and ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Bernnat, Wolfgang; Buck, Michael [Stuttgart Univ. (DE). Inst. fuer Kernenergetik und Energiesysteme (IKE); Pasichnyk, Ihor; Zwermann, Winfried [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Garching (Germany)

    2011-07-01

    The analysis of pin wise power distribution in LWR lattices under realistic thermal hydraulics conditions can be performed with coupled neutronics and thermal hydraulics codes. Due to today's availability of powerful parallel computer resources for the neutronics part Monte Carlo codes can be applied with the advantage that no homogenization and energy group approximations has to be used. Realistic operational conditions require that for all pins the material and temperature distributions must be taken into account. Using Monte Carlo codes like MCNP5 a very large number of input zones with different material composition or temperature must be specified. The corresponding thermal hydraulics data - moderator temperatures, densities and fuel temperatures - must be calculated by means of an appropriate thermal hydraulics code. This paper describes the coupling of MCNP5 with the system code ATHLET for the analysis of the detailed power distribution in a PWR. The PWR reactor model was simplified according to the stationary part of the Purdue benchmark. This means that the burnup of all assemblies is radially and axially constant and the temperature in an assembly varies only axially. The nuclide compositions of the different assemblies (UOX and MOX) for specified burnups were taken from the benchmark specification. (orig.)

  3. HTGR reactor physics, thermal-hydraulics and depletion uncertainty analysis: a proposed IAEA coordinated research project

    International Nuclear Information System (INIS)

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis and uncertainty analysis methods. In order to benefit from recent advances in modeling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Uncertainty and sensitivity studies are an essential component of any significant effort in data and simulation improvement. In February 2009, the Technical Working Group on Gas-Cooled Reactors recommended that the proposed IAEA Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modeling be implemented. In the paper the current status and plan are presented. The CRP will also benefit from interactions with the currently ongoing OECD/NEA Light Water Reactor (LWR) UAM benchmark activity by taking into consideration the peculiarities of HTGR designs and simulation requirements. (author)

  4. In-core thermal hydraulic and fission product calculations for severe fuel damage analyses

    International Nuclear Information System (INIS)

    Best-estimate calculations of realistic source terms are presented which reduce uncertainties in predicting fission product release from the UO2 fuel over the temperature range between 770 K and 3000 K. The proposed method of correlation includes such fuel morphology effects as equiaxed fuel grain growth and fuel-cladding interaction. The method correlates the product of fuel release rate and equiaxed grain size with the inverse fuel temperature to yield a bulk mass transfer correlation. It was found that less and slower releases are predicted utilizing the bulk mass transfer correlation than such conventional literature correlations as the steam oxidation model and the NUREG-0956 correlation. A Severe Fuel Damage (SFD) analysis code was developed to perform the thermal hydraulic and fission product calculations needed to analyze the Power Burst Facility SFD tests. The predictions utilizing the bulk mass transfer correlations overall followed the experimental time-release histories during the course of the heatup, power hold and cooldown phases of the transients. Good agreements were achieved for the integral releases both in timing and in magnitude. The proposed bulk mass transfer correlations can be applied to both current and advanced light water reactor fuels. 17 refs., 8 figs., 3 tabs

  5. Computational models for thermal-hydraulic assessment of TADSEA and its use for hydrogen production

    International Nuclear Information System (INIS)

    The Transmutation Advanced Device for Sustainable Energy Applications (TADSEA) is a pebble-bed Accelerator Driven System (ADS) with a graphite-gas configuration, designed for nuclear waste transmutation and for obtaining heat at very high temperatures to produce hydrogen. In previous work, the TADSEA's nuclear core was considered as a porous medium performed with a CFD code and thermal-hydraulic studies of the nuclear core were presented. In this paper, three critical fuel elements groups were defined regarding their position inside the core. In this article, the heat transfer from the fuel to the coolant was analyzed for the three core states during normal operation. The heat transfer inside the spherical fuel elements was also studied with a realistic CFD model of the critical elements groups. During the steady state, no critical elements reached the limit temperature of this type of fuel. Also, it is presented a model built in ANSYS for the simulation and optimization of high- temperature electrolysis using the TADSEA as a heat source. A flow diagram of the electrolysis process with the high temperature electrolyzer as the main component using TADSEA as an energy source is finally proposed and discussed. (Author)

  6. Elements of validation for LWRs thermal hydraulic studies with FLICA-OVAP

    International Nuclear Information System (INIS)

    FLICA-OVAP is an advanced two-phase flow thermal-hydraulics code based on a full 3D subchannel approach. It is designed to analyze flows in Light Water Reactors (LWRs) cores such as PWRs, BWRs and experimental reactors. Therefore its applicability covers all ranges of operating conditions for water-cooled reactors. This paper presents an overview of FLICA-OVAP modeling capabilities for applications in nuclear reactors design and safety analysis. A validation matrix is proposed and its results are presented. The matrix covers a wide range of selected phenomena, which are relevant for thermalhydraulics studies. Therefore the different FLICA-OVAP physical correlations addressed in the current study include single phase and two-phase friction factors, single phase and boiling heat transfer, turbulence and critical heat flux. Results of the FLICA-OVAP validation studies highlight the capabilities of the code to well-predict two-phase flows in Light Water Reactors for both normal operation and under accidental circumstances. Future developments as well as validation activities are also summarized. (author)

  7. Thermal-hydraulic limitations on water-cooled limiters

    International Nuclear Information System (INIS)

    An assessment of the cooling requirements for fusion reactor components, such as the first wall and limiter/divertor, was carried out using pressurized water as the coolant. In order to establish the coolant operating conditions, a survey of the literature on departure from nucleate boiling, critical heat flux, asymmetrical heating and heat transfer augmentation techniques was carried out. The experimental data and the empirical correlations indicate that thermal protection for the fusion reactor components based on current design concepts can be provided with an adequate margin of safety without resorting to either high coolant velocities, excessive coolant pressures, or heat transfer augmentation techniques. If, however, the future designs require heat transfer enhancement techniques, experimental verification would be necessary since no data on heat transfer augmentation techniques exist for complex geometries, especially under asymmetrically heated conditions. Since the data presented herein concern primarily thermal protection, the final design should consider other factors such as thermal stresses, temperature limits, and fatigue

  8. Advanced hydraulic fracturing methods to create in situ reactive barriers

    Energy Technology Data Exchange (ETDEWEB)

    Murdoch, L. [FRx Inc., Cincinnati, OH (United States)]|[Clemson Univ., SC (United States); Siegrist, B. [Oak Ridge National Lab., TN (United States); Vesper, S. [Univ. of Cincinnati, OH (United States)] [and others

    1997-12-31

    Many contaminated areas consist of a source area and a plume. In the source area, the contaminant moves vertically downward from a release point through the vadose zone to an underlying saturated region. Where contaminants are organic liquids, NAPL may accumulate on the water table, or it may continue to migrate downward through the saturated region. Early developments of permeable barrier technology have focused on intercepting horizontally moving plumes with vertical structures, such as trenches, filled with reactive material capable of immobilizing or degrading dissolved contaminants. This focus resulted in part from a need to economically treat the potentially large volumes of contaminated water in a plume, and in part from the availability of construction technology to create the vertical structures that could house reactive compounds. Contaminant source areas, however, have thus far remained largely excluded from the application of permeable barrier technology. One reason for this is the lack of conventional construction methods for creating suitable horizontal structures that would place reactive materials in the path of downward-moving contaminants. Methods of hydraulic fracturing have been widely used to create flat-lying to gently dipping layers of granular material in unconsolidated sediments. Most applications thus far have involved filling fractures with coarse-grained sand to create permeable layers that will increase the discharge of wells recovering contaminated water or vapor. However, it is possible to fill fractures with other compounds that alter the chemical composition of the subsurface. One early application involved development and field testing micro-encapsulated sodium percarbonate, a solid compound that releases oxygen and can create aerobic conditions suitable for biodegradation in the subsurface for several months.

  9. Advanced hydraulic fracturing methods to create in situ reactive barriers

    International Nuclear Information System (INIS)

    Many contaminated areas consist of a source area and a plume. In the source area, the contaminant moves vertically downward from a release point through the vadose zone to an underlying saturated region. Where contaminants are organic liquids, NAPL may accumulate on the water table, or it may continue to migrate downward through the saturated region. Early developments of permeable barrier technology have focused on intercepting horizontally moving plumes with vertical structures, such as trenches, filled with reactive material capable of immobilizing or degrading dissolved contaminants. This focus resulted in part from a need to economically treat the potentially large volumes of contaminated water in a plume, and in part from the availability of construction technology to create the vertical structures that could house reactive compounds. Contaminant source areas, however, have thus far remained largely excluded from the application of permeable barrier technology. One reason for this is the lack of conventional construction methods for creating suitable horizontal structures that would place reactive materials in the path of downward-moving contaminants. Methods of hydraulic fracturing have been widely used to create flat-lying to gently dipping layers of granular material in unconsolidated sediments. Most applications thus far have involved filling fractures with coarse-grained sand to create permeable layers that will increase the discharge of wells recovering contaminated water or vapor. However, it is possible to fill fractures with other compounds that alter the chemical composition of the subsurface. One early application involved development and field testing micro-encapsulated sodium percarbonate, a solid compound that releases oxygen and can create aerobic conditions suitable for biodegradation in the subsurface for several months

  10. Nuclear Thermal Propulsion for Advanced Space Exploration

    Science.gov (United States)

    Houts, M. G.; Borowski, S. K.; George, J. A.; Kim, T.; Emrich, W. J.; Hickman, R. R.; Broadway, J. W.; Gerrish, H. P.; Adams, R. B.

    2012-01-01

    The fundamental capability of Nuclear Thermal Propulsion (NTP) is game changing for space exploration. A first generation Nuclear Cryogenic Propulsion Stage (NCPS) based on NTP could provide high thrust at a specific impulse above 900 s, roughly double that of state of the art chemical engines. Characteristics of fission and NTP indicate that useful first generation systems will provide a foundation for future systems with extremely high performance. The role of the NCPS in the development of advanced nuclear propulsion systems could be analogous to the role of the DC-3 in the development of advanced aviation. Progress made under the NCPS project could help enable both advanced NTP and advanced Nuclear Electric Propulsion (NEP).

  11. Shape optimization of a printed-circuit heat exchanger to enhance thermal-hydraulic performance

    International Nuclear Information System (INIS)

    Printed circuit heat exchanger (PCHE) is recently considered as a recuperator for the high temperature gas cooled reactor. In this work, the zigzag-channels of a PCHE have been optimized by using three-dimensional Reynolds-Averaged Navier-Stokes (RANS) analysis and response surface approximation (RSA) modeling technique to enhance thermal-hydraulic performance. Shear stress transport turbulence model is used as a turbulence closure. The objective function is defined as a linear combination of the functions related to heat transfer and friction loss of the PCHE, respectively. Three geometric design variables viz., the ratio of the radius of the fillet to hydraulic diameter of the channels, the ratio of wavelength to hydraulic diameter of the channels, and the ratio of wave height to hydraulic diameter of the channels, are used for the optimization. Design points are selected through Latin-hypercube sampling. The optimal design is determined through the RSA model which uses RANS derived calculations at the design points. The results show that the optimum shape enhances considerably the thermal-hydraulic performance than a reference shape. (authors)

  12. Thermal-hydraulic characteristics and performance of 3D wavy channel based printed circuit heat exchanger

    International Nuclear Information System (INIS)

    CFD study is done here to propose an efficient PCHE (Printed Circuit Heat Exchanger) model; used as a recuperator in International Thermonuclear Experimental Reactor (ITER). 3D steady state conjugate heat-transfer numerical simulations are done; considering the variation of thermo-physical properties as a function of temperature. Helium is used as a working fluid and alloy 617 as solid substrate. The study is done for various angle of bend (θ = 0°(straight), 5°, 10° and 15°) and Reynolds number (Re = 350, 700, 1400 and 2100). Various types of flow patterns, within one wavy-section, are presented to analyze thermal-hydraulic characteristics. Thermal hydraulic performance parameters are presented for the various wavy-sections as well as within a section; and for the complete PCHE model. Heat transfer enhancement as compared to pressure penalty is higher for the wavy channel; and increases with increasing Re and θ. Wavy as compared to plane channel based PCHE is demonstrated here to give better thermal-hydraulic performance. A detailed characteristics as well as performance-parameters for thermal hydraulics in a 3D wavy channel based PCHE model − not found in the literature − is presented here. - Highlights: • Studied effect of Reynolds number and angle of bend. • Analyzed thermal-hydraulic characteristics, by various types of flow pat-terns. • Demonstrated an increase in local heat flux due to change in the flow-direction. • Demonstrated better performance of wavy as compared to plane channel based PCHE. • Proposed correlation for friction factor and Nusselt number

  13. Coupled fully 3D neutron kinetics thermal-hydraulic computations for DNB analysis on PWRs

    International Nuclear Information System (INIS)

    Departure from Nucleate Boiling (DNB) is one of the major limiting factors of Pressurized Water Reactors (PWRs). Safety requires that occurrence of DNB should be precluded under normal or incidental operating conditions. To perform Main Steam Line Break (MSLB) accident calculations EDF have developed its own numerical tool OSCARD based on: the thermal-hydraulic THYC code for DNB analysis, the neutron kinetics COCCINELLE code for power distribution computations, the thermal-hydraulic CATHARE code to provide boundary conditions analysis with system scale computation. With OSCARD a fully three-dimensional (3D) representation of the core is proposed in conjunction with a two-phase flow porous-body approach (THYC) and two-group diffusion equations in the axial and lateral directions with Doppler and void reactivity feedback effects (COCCINELLE). OSCARD provides EDF with an alternative and independent way of evaluating fuel performance and safety margins. In the licensed approach, the coupled 3D neutron kinetics and thermal-hydraulic part of OSCARD steady computations is used to produce 3D power distribution in the reactor core at the most penalizing moment of the transient. Then this distribution is used as an input for THYC to perform thermal-hydraulic subchannel analysis. This 3 steps approach is used with simple conservative and bounding analysis assumptions, that can not occur in reality. In a prospective approach, OSCARD enables to combine thermal-hydraulic subchannel analysis with the neutron kinetics radial average channel model using a nodalization of one quarter of fuel assembly in order to perform one step DNB analysis. (author)

  14. The module CCM for the simulation of the thermal-hydraulic situation within a coolant channel

    International Nuclear Information System (INIS)

    A coolant channel module (Cc) will be presented which aim is to simulate, in a very general way, the thermal-hydraulic behaviour of single- and two-phase fluids flowing along a heated (or cooled) vertical, inclined or horizontal coolant channel. It is based on a theoretical drift-flux supported 3-equation mixture-fluid model describing the steady state and transient behaviour of characteristic thermal-hydraulic parameters of a single- and two-phase flow within such a channel. The module can be applied as an element within an overall theoretical model for large and complex plant assemblies (PWR and BWR core channels, parallel channels in 3D cores, primary and secondary sides of different steam generators types etc.). The model refers to a general (basic) coolant channel (BC) which can consists of different flow regimes. The BC has thus to be subdivided accordingly into a number of subchannels (SC-s). All of them can belong, however, to only two types of SC-s (single-phase fluid with subcooled water or superheated steam or a two-phase flow regime). For both of them the possibility of variable entrance or outlet positions has to be considered. For discretization purposes the BC (and thus also the SC-s) have to be subdivided into a number of (BC and SC) nodes, discretizing thus the conservation equations for mass, energy and momentum along these nodes by applying a very general spatial procedure, namely a 'modified finite volume method'. A special quadratic polygon approximation method (PAX procedure) helps then to establish a connection between nodal boundary and mean nodal parameters. Considering their constitutive equations (among them an adequate drift-flux correlation package) yields finally a set of non-linear algebraic and non-linear ordinary differential equations for the characteristic parameters of each of these SC nodes (mass flow, pressure drop, coolant temperature and/or void fraction). Based on this theory a code package (CCM) could be established

  15. Thermal Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Water: Performance and Stability

    Science.gov (United States)

    Lisowski, Darius D.

    This experimental study investigated the thermal hydraulic behavior and boiling mechanisms present in a scaled reactor cavity cooling system (RCCS). The experimental facility reflects a ¼ scale model of one conceptual design for decay heat removal in advanced GenIV nuclear reactors. Radiant heaters supply up to 25 kW/m2 onto a three parallel riser tube and cooling panel test section assembly, representative of a 5° sector model of the full scale concept. Derived similarity relations have preserved the thermal hydraulic flow patterns and integral system response, ensuring relevant data and similarity among scales. Attention will first be given to the characterization of design features, form and heat losses, nominal behavior, repeatability, and data uncertainty. Then, tests performed in single-phase have evaluated the steady-state behavior. Following, the transition to saturation and subsequent boiling allowed investigations onto four parametric effects at two-phase flow and will be the primary focus area of remaining analysis. Baseline conditions at two-phase flow were defined by 15.19 kW of heated power and 80% coolant inventory, and resulted in semi-periodic system oscillations by the mechanism of hydrostatic head fluctuations. Void generation was the result of adiabatic expansion of the fluid due to a reduction in hydrostatic head pressure, a phenomena similar to flashing. At higher powers of 17.84 and 20.49 kW, this effect was augmented, creating large flow excursions that followed a smooth and sinusoidal shaped path. Stabilization can occur if the steam outflow condition incorporates a nominal restriction, as it will serve to buffer the short time scale excursions of the gas space pressure and dampen oscillations. The influences of an inlet restriction, imposed by an orifice plate, introduced subcooling boiling within the heated core and resulted in chaotic interactions among the parallel risers. The penultimate parametric examined effects of boil-off and

  16. Advancement of experimentation for measuring hydraulic conductivity of bentonite using high-pressure consolidation test apparatus

    International Nuclear Information System (INIS)

    In the geological disposal facility of high-level radioactive wastes, it is important to grasp the hydraulic conductivity characteristic of bentonite. The purpose of this study is the advancement of the examination method for the measurement of a more reliable hydraulic conductivity using high-pressure consolidation test apparatus (maximum consolidation pressure 10MPa). Consequently, it succeeded in improving the reliability of data by raising the resolution of displacement used for an examination, increasing to 80 the number of measurement data for 2 minutes after making each consolidation pressure act on the occasion of measurement and adopting the data of a high consolidation pressure (more than 5.88MPa) stage. (author)

  17. Thermal-Hydraulic Feedback Module for BGCore System

    International Nuclear Information System (INIS)

    The need for accurate fuel management modeling in Generation IV (LFR, HTGR, etc) reactors has motivated the development of a new and comprehensive code (the BGCore system) for core analysis of advanced reactors. This effort is justified since there are currently no high fidelity codes, which are capable of performing for different types of advanced reactors calculations. Reliable modeling of the core performance requires an adequate modeling of a wide range of physical processes, such as fuel depletion and the temperature distribution in the main core components

  18. Neutronic and thermal-hydraulic coupling for 3D reactor core modeling combining MCB and fluent

    Directory of Open Access Journals (Sweden)

    Królikowski Igor P.

    2015-09-01

    Full Text Available Three-dimensional simulations of neutronics and thermal hydraulics of nuclear reactors are a tool used to design nuclear reactors. The coupling of MCB and FLUENT is presented, MCB allows to simulate neutronics, whereas FLUENT is computational fluid dynamics (CFD code. The main purpose of the coupling is to exchange data such as temperature and power profile between both codes. Temperature required as an input parameter for neutronics is significant since cross sections of nuclear reactions depend on temperature. Temperature may be calculated in thermal hydraulics, but this analysis needs as an input the power profile, which is a result from neutronic simulations. Exchange of data between both analyses is required to solve this problem. The coupling is a better solution compared to the assumption of estimated values of the temperatures or the power profiles; therefore the coupled analysis was created. This analysis includes single transient neutronic simulation and several steady-state thermal simulations. The power profile is generated in defined points in time during the neutronic simulation for the thermal analysis to calculate temperature. The coupled simulation gives information about thermal behavior of the reactor, nuclear reactions in the core, and the fuel evolution in time. Results show that there is strong influence of neutronics on thermal hydraulics. This impact is stronger than the impact of thermal hydraulics on neutronics. Influence of the coupling on temperature and neutron multiplication factor is presented. The analysis has been performed for the ELECTRA reactor, which is lead-cooled fast reactor concept, where the coolant fl ow is generated only by natural convection

  19. Thermal hydraulics and mechanics research on fusion blanket system

    International Nuclear Information System (INIS)

    In-vessel components such as Blanket and Divertor in a fusion reactor have a function of exhausting high heat and particle loads in order to maintain the structural soundness of the reactor. In the International Thermonuclear Experimental Reactor called ITER, build by ITER Organization under the framework of collaboration of seven parties including Japan, there are two kinds of blanket systems will be install. One is a shield blanket, which consists of a first wall (FW) and a block module shielding against neutron flux to a vacuum chamber and a superconducting magnet system. The other blanket system is called as a Test Blanket Module (TBM). TBM is a kind of prototype blanket for a fusion power plant and has functions of breeding of tritium (T) and extraction of energy from fusion plasma. TBM consists of FW and T-breeding / neutron (n)-multiplier zone. A concept of TBM developed by JAEA is water-cooled pebble-bed type, which means that FW and other structures are cooled by pressurized high temperature water and T-breeding / n-multiplier zone consists of multiple layers of pebble bed made of T-breeding and n-multiplier material. This paper describes the status of R and Ds on FW and pebble beds from the view of thermo-hydraulics and mechanics. (author)

  20. Investigation of coolant thermal mixing within 28-element CANDU fuel bundles using the ASSERT-PV thermal hydraulics code

    International Nuclear Information System (INIS)

    This paper presents the results of a study of the thermal mixing of single-phase coolant in 28-element CANDU fuel bundles under steady-state conditions. The study, which is based on simulations performed using the ASSERT-PV thermal hydraulic code, consists of two main parts. In the first part the various physical mechanisms that contribute to coolant mixing are identified and their impact is isolated via ASSERT-PV simulations. The second part is concerned with development of a preliminary model suitable for use in the fuel and fuel channel code FACTAR to predict the thermal mixing that occurs between flow annuli. (author)

  1. Thermal APU/hydraulics analysis program. User's guide and programmer's manual

    Science.gov (United States)

    Deluna, T. A.

    1976-01-01

    The User's Guide information plus program description necessary to run and have a general understanding of the Thermal APU/Hydraulics Analysis Program (TAHAP) is described. This information consists of general descriptions of the APU/hydraulic system and the TAHAP model, input and output data descriptions, and specific subroutine requirements. Deck setups and input data formats are included and other necessary and/or helpful information for using TAHAP is given. The math model descriptions for the driver program and each of its supporting subroutines are outlined.

  2. Simulation of condensation in the presence of noncondensible gases using the French thermal hydraulic code CATHARE

    International Nuclear Information System (INIS)

    The advanced designs of new light-water-reactors have provided new areas in which condensation is an important model of heat transfer. In an application where condensation heat transfer is important, the proposed advanced passive boiling water reactor, the simplified boiling water reactor (SBWR), design utilizes as a mean component of the passive containment cooling system the isolation condenser (IC). The IC has as its main function to provide a passive heat exchanger for the removal of the reactor primary coolant energy and that of core decay heat energy. The energy from the fluid stream is deposited in large reservoirs situated above the core within the containment structure. In order for the IC to perform the task of energy removal, the IC must be able to remove sufficient energy from the fluid stream such that the design pressure of the containment is not exceeded, and that the containment pressure is reduced overall in the long run. After a typical postulated loss-of-coolant accident, a steam/air mixture from the reactor pressure vessel may flow to the IC. Once at the IC, the steam/air mixture condenses in downward flow through a bundle of tubes, and the condensed steam drains back to the reactor pressure vessel. In this investigation, a simulation of the University of California, Berkeley, Condensation Studies was performed using the CATHARE V1.3 U-1 thermal hydraulic computer code to model the phenomena of condensation in a vertical test section. Bulk steam/gas temperature, condensing tube heat flux, and local overall heat transfer coefficients compare well with that of the experiment. Cooling jacket bulk temperature profiles were also evaluated

  3. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  4. Thermal-hydraulic analysis of a 600 MW supercritical CFB boiler with low mass flux

    International Nuclear Information System (INIS)

    Supercritical Circulating Fluidized Bed (CFB) boiler becomes an important development trend for coal-fired power plant and thermal-hydraulic analysis is a key factor for the design and operation of water wall. According to the boiler structure and furnace-sided heat flux, the water wall system of a 600 MW supercritical CFB boiler is treated in this paper as a flow network consisting of series-parallel loops, pressure grids and connecting tubes. A mathematical model for predicting the thermal-hydraulic characteristics in boiler heating surface is based on the mass, momentum and energy conservation equations of these components, which introduces numerous empirical correlations available for heat transfer and hydraulic resistance calculation. Mass flux distribution and pressure drop data in the water wall at 30%, 75% and 100% of the boiler maximum continuous rating (BMCR) are obtained by iteratively solving the model. Simultaneity, outlet vapor temperatures and metal temperatures in water wall tubes are estimated. The results show good heat transfer performance and low flow resistance, which implies that the water wall design of supercritical CFB boiler is applicable. - Highlights: → We proposed a model for thermal-hydraulic analysis of boiler heating surface. → The model is applied in a 600 MW supercritical CFB boiler. → We explore the pressure drop, mass flux and temperature distribution in water wall. → The operating safety of boiler is estimated. → The results show good heat transfer performance and low flow resistance.

  5. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    International Nuclear Information System (INIS)

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations

  6. Design of a Resistively Heated Thermal Hydraulic Simulator for Nuclear Rocket Reactor Cores

    Science.gov (United States)

    Litchford, Ron J.; Foote, John P.; Ramachandran, Narayanan; Wang, Ten-See; Anghaie, Samim

    2007-01-01

    A preliminary design study is presented for a non-nuclear test facility which uses ohmic heating to replicate the thermal hydraulic characteristics of solid core nuclear reactor fuel element passages. The basis for this testing capability is a recently commissioned nuclear thermal rocket environments simulator, which uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce high-temperature pressurized hydrogen flows representative of reactor core environments, excepting radiation effects. Initially, the baseline test fixture for this non-nuclear environments simulator was configured for long duration hot hydrogen exposure of small cylindrical material specimens as a low cost means of evaluating material compatibility. It became evident, however, that additional functionality enhancements were needed to permit a critical examination of thermal hydraulic effects in fuel element passages. Thus, a design configuration was conceived whereby a short tubular material specimen, representing a fuel element passage segment, is surrounded by a backside resistive tungsten heater element and mounted within a self-contained module that inserts directly into the baseline test fixture assembly. With this configuration, it becomes possible to create an inward directed radial thermal gradient within the tubular material specimen such that the wall-to-gas heat flux characteristics of a typical fuel element passage are effectively simulated. The results of a preliminary engineering study for this innovative concept are fully summarized, including high-fidelity multi-physics thermal hydraulic simulations and detailed design features.

  7. Thermal-hydraulic calculations for KUHFR with reduced enrichment uranium fuel

    International Nuclear Information System (INIS)

    This report provides the preliminary results of the thermal-hydraulic calculations to study the safety aspects in fueling the KUHFR with reduced enrichment uranium. The calculations were based on what was outlined in the Safety Analysis Report for the KUHFR and the guidebook for research reactor core conversion, IAEA-TECDOC-233, published by the International Atomic Energy Agency. No significant differences in the thermal-hydraulic operating conditions have been found between HEU and MEU fuels. However, in LEU cases, the combination of three factors - larger power peaking with LEU fuel, smaller thermal conductivity of U3O8-Al fuel with high uranium densities, and thicker fuel meat - resulted in higher maximum fuel and surface temperatures with the LEU oxide fuel. (author)

  8. ASFRE: a computer code for single-phase subchannel thermal hydraulic analysis of LMFBR single subassembly

    International Nuclear Information System (INIS)

    The objectives of this work is to develop a computer code ASFRE which analyzes 3D-thermo-hydraulic behaviors of coolant and fuel pins in an LMFBR subassembly under accident conditions such as the local blockage, loss of flow and transient over power accident conditions. Analytical models, calculation procedures and sample calculations for typical experiments are described. The ASFRE code consists of two parts, namely coolant calculation part and fuel pin calculation. The coolant thermal-hydraulic analysis employs basically subchannel analysis approach and the program solves transient mass, momentum and energy conservation equations. The fuel pin thermal analysis program solves transient heat conduction equations by finite difference method in cylindrical coordinate system. Fuel temperature distribution and thermal expansion are calculated taking into account of intra/inter-pin-flux-depression and fuel restructuring. And wire wrap spacer effects for coolant behavior and heat loss through the wrapper tube are also simulated. (author)

  9. Estimating soil moisture and soil thermal and hydraulic properties by assimilating soil temperatures using a particle batch smoother

    Science.gov (United States)

    Dong, Jianzhi; Steele-Dunne, Susan C.; Ochsner, Tyson E.; Giesen, Nick van de

    2016-05-01

    This study investigates the potential of estimating the soil moisture profile and the soil thermal and hydraulic properties by assimilating soil temperature at shallow depths using a particle batch smoother (PBS) using synthetic tests. Soil hydraulic properties influence the redistribution of soil moisture within the soil profile. Soil moisture, in turn, influences the soil thermal properties and surface energy balance through evaporation, and hence the soil heat transfer. Synthetic experiments were used to test the hypothesis that assimilating soil temperature observations could lead to improved estimates of soil hydraulic properties. We also compared different data assimilation strategies to investigate the added value of jointly estimating soil thermal and hydraulic properties in soil moisture profile estimation. Results show that both soil thermal and hydraulic properties can be estimated using shallow soil temperatures. Jointly updating soil hydraulic properties and soil states yields robust and accurate soil moisture estimates. Further improvement is observed when soil thermal properties were also estimated together with the soil hydraulic properties and soil states. Finally, we show that the inclusion of a tuning factor to prevent rapid fluctuations of parameter estimation, yields improved soil moisture, temperature, and thermal and hydraulic properties.

  10. FLUENT-based neutronics and thermal-hydraulics coupling calculation for a liquid-fuel molten salt reactor

    International Nuclear Information System (INIS)

    Molten Salt Reactor (MSR) is the only one using liquid fuel in the six candidate reactors of the Generation IV advanced nuclear power systems with expected remarkable advantages in safety, economics, sustainability, and proliferation resistance. The strong coupling between neutronics and thermal-hydraulics due to fuel movement in the liquid-fuel MSRs induces many new challenges in reactor analyses from the perspective of both theoretical models and solution methods. In this study, the multi-group diffusion theory was adopted to deduce the neutronics model for the liquid-fuel MSRs, in which the salt flow effects on the delayed neutron precursor distributions in space were considered particularly. Since the liquid-fuel salt is a Newton fluid, the single-phase thermal hydraulics model for liquid-fuel MSRs was generally established based on the fundamental laws of the mass, momentum and energy conservation equations as used in the computational fluid dynamic (CFD) method. Since the control equations of the neutronic model can be written in the same form of those solved in the CFD softwares, a neutronics and thermal-hydraulics coupling scheme was proposed and a program was developed based on the FLUENT software by using its user-defined functions and subroutines (UDF and UDS). This program was applied to perform the steady state calculation of the molten salt fast reactor (MSFR), and the main results such as the space distributions of the neutron fluxes, delayed neutron precursors, temperatures, velocities were obtained. The results show that the liquid fuel flow influences the delayed neutron precursors significantly, while slightly affects the neutron fluxes. The flow in the MSFR core generates a vortex near the fertile tank leading the maximal temperature to about 1200 K at the centre of the vortex, which will be optimized in the future core design. (author)

  11. Advanced thermally assisted surface engineering processes

    CERN Document Server

    Chattopadhyay, Ramnarayan

    2007-01-01

    Preface. Acknowledgements. 1: Wear, Surface Heat and Surface Engineering. 2: Plasma Assisted Thermal Processes. 3: Ion Beam Processes. 4: Electron Beam Processes. 5: Microwave Assisted Surface Modification Processes. 6: Laser Assisted Surface Engineering Processes. 7: Solar Energy for Surface Modifications. 8: Combustion Processes for Surface Modification. 9: Friction Weld Surfacing. 10: Induction Surface Modification Processes. 11: Surfacing by Spark Deposition Processes. 12: Arc Assisted Advanced Surface Engineering Processes. 13: Hot Isostatic Press. 14: Fluid Bed Processes. 15: P

  12. Effects of substrate properties on the hydraulic and thermal behavior of a green roof

    Science.gov (United States)

    Sandoval, V. P.; Suarez, F. I.; Victorero, F.; Bonilla, C.; Gironas, J. A.; Vera, S.; Bustamante, W.; Rojas, V.; Pasten, P.

    2014-12-01

    Green roofs are a sustainable urban development solution that incorporates a growing media (also known as substrate) and vegetation into infrastructures to reach additional benefits such as the reduction of: rooftop runoff peak flows, roof surface temperatures, energy utilized for cooling/heating buildings, and the heat island effect. The substrate is a key component of the green roof that allows achieving these benefits. It is an artificial soil that has an improved behavior compared to natural soils, facilitating vegetation growth, water storage and typically with smaller densities to reduce the loads over the structures. Therefore, it is important to study the effects of substrate properties on green roof performance. The objective of this study is to investigate the physical properties of four substrates designed to improve the behavior of a green roof, and to study their impact on the efficiency of a green roof. The substrates that were investigated are: organic soil; crushed bricks; a mixture of mineral soil with perlite; and a mixture of crushed bricks and organic soil. The thermal properties (thermal conductivity, volumetric heat capacity and thermal diffusivity) were measured using a dual needle probe (Decagon Devices, Inc.) at different saturation levels, and the hydraulic properties were measured with a constant head permeameter (hydraulic conductivity) and a pressure plate extractor (water retention curve). This characterization, combined with numerical models, allows understanding the effect of these properties on the hydraulic and thermal behavior of a green roof. Results show that substrates composed by crushed bricks improve the thermal insulation of infrastructures and at the same time, retain more water in their pores. Simulation results also show that the hydraulic and thermal behavior of a green roof strongly depends on the moisture content prior to a rainstorm.

  13. Development of a coupled neutronic/thermal-hydraulic tool with multi-scale capabilities and applications to HPLWR core analysis

    International Nuclear Information System (INIS)

    Highlights: → Advanced analysis and design techniques for innovative reactors are addressed. → Detailed investigation of a 3 pass core design with a multi-physics-scales tool. → Coupled 40-group neutron transport/equivalent channels TH core analyses methods. → Multi-scale capabilities: from equivalent channels to sub-channel pin-by-pin study. → High fidelity approach: reduction of conservatism involved in core simulations. - Abstract: The High Performance Light Water Reactor (HPLWR) is a thermal spectrum nuclear reactor cooled and moderated with light water operated at supercritical pressure. It is an innovative reactor concept, which requires developing and applying advanced analysis tools as described in the paper. The relevant water density reduction associated with the heat-up, together with the multi-pass core design, results in a pronounced coupling between neutronic and thermal-hydraulic analyses, which takes into account the strong natural influence of the in-core distribution of power generation and water properties. The neutron flux gradients within the multi-pass core, together with the pronounced dependence of water properties on the temperature, require to consider a fine spatial resolution in which the individual fuel pins are resolved to provide precise evaluation of the clad temperature, currently considered as one of the crucial design criteria. These goals have been achieved considering an advanced analysis method based on the usage of existing codes which have been coupled with developed interfaces. Initially neutronic and thermal-hydraulic full core calculations have been iterated until a consistent solution is found to determine the steady state full power condition of the HPLWR core. Results of few group neutronic analyses might be less reliable in case of HPLWR 3-pass core than for conventional LWRs because of considerable changes of the neutron spectrum within the core, hence 40 groups transport theory has been preferred to the

  14. AP600 design certification thermal hydraulics testing and analysis

    International Nuclear Information System (INIS)

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

  15. AP600 design certification thermal hydraulics testing and analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  16. Rapid Thermal-Hydraulic Analysis and Design Optimization of ITER Upper ELM Coils

    International Nuclear Information System (INIS)

    ITER edge localized mode (ELM) coils are important components of the in-vessel coils (IVCs) and they are designed for mitigating or suppressing ELMs. The coils located on the vacuum vessel (VV) and behind the blanket are subjected to high temperature due to the nuclear heat from the plasma, the Ohmic heat induced by the working current and the thermal radiation from the environment. The water serves as coolant to remove the heat deposited into the coils. Based on the results of nuclear analysis, the thermal-hydraulic analysis is performed for the preliminary design of upper ELM coils using a rapid evaluation method based on 1D treatment. The thermal-hydraulic design and operating parameters including the water flow velocity are optimized. It is found that the rapid evaluation method based on 1D treatment is feasible and reliable. According to the rapid analysis method, the thermal hydraulic parameters of two water flow schemes are computed and proved similar to each other, providing an effective basis for the coil design. Finally, considering jointly the pressure drop requirement and the cooling capacity, the flow velocity is optimized to 5 m/s. (fusion engineering)

  17. The 25 kWe solar thermal Stirling hydraulic engine system: Conceptual design

    Science.gov (United States)

    White, Maurice; Emigh, Grant; Noble, Jack; Riggle, Peter; Sorenson, Torvald

    1988-01-01

    The conceptual design and analysis of a solar thermal free-piston Stirling hydraulic engine system designed to deliver 25 kWe when coupled to a 11 meter test bed concentrator is documented. A manufacturing cost assessment for 10,000 units per year was made. The design meets all program objectives including a 60,000 hr design life, dynamic balancing, fully automated control, more than 33.3 percent overall system efficiency, properly conditioned power, maximum utilization of annualized insolation, and projected production costs. The system incorporates a simple, rugged, reliable pool boiler reflux heat pipe to transfer heat from the solar receiver to the Stirling engine. The free-piston engine produces high pressure hydraulic flow which powers a commercial hydraulic motor that, in turn, drives a commercial rotary induction generator. The Stirling hydraulic engine uses hermetic bellows seals to separate helium working gas from hydraulic fluid which provides hydrodynamic lubrication to all moving parts. Maximum utilization of highly refined, field proven commercial components for electric power generation minimizes development cost and risk.

  18. Thermal-hydraulics of a steam discharge system

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Yoon Sub; Yoo, Keun Jong; Kim, Yun Sik; Lee, Ki Yung; Wooi Myung Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-12-01

    Physical characteristics of the steam discharge system of PWR, which consists of valves, piping, steam sparger, and steam accommodating tank, have been analyzed and the analysis system has been set up for development of the analysis and design methodology for the system. The work was the results of the final year research in the planned research period of three years. Analysis has been made for the characteristics of rapid transient and steady flow in the piping, bubble behavior and wall pressure oscillation in a large and open tank, steam condensation, thermal mixing performance in a pool, and small and hermetically sealed tank performance. Based on the analysis results, experimental requirements for the development provided. Finally, for the further work in developing KNGR, the implementing approach related to this work has been purposed. 42 figs, 2 tabs, 29 refs. (Author).

  19. The use of software quality assurance techniques in the validation of results in nuclear thermal-hydraulic safety research

    International Nuclear Information System (INIS)

    The Nuclear Engineering Department at Israel Electric Company has been engaged for a number of years in a joint research agreement with the Technion Nuclear Reactor Research Group on various thermal-hydraulic aspects of reactor design and safety. Besides developing their own analytical models, the researchers rely heavily on the RELAPS computer code in their analyses. The RELAPS series are general purpose, thermal-hydraulic system codes, used to simulate system response (such as the RCS) to transients and accidents. They are based on solving the equations of conservation of mass, energy and momentum within the system being modeled, where the model is a series of control volumes connected by junctions. The equations are solved simultaneously in each volume and junction using a finite difference numerical scheme. As an example, a recent report refers to containment response to a large LOCA in an AP600-like advanced rector. This work has been performed using RELAPS/Mod2. Accidents like LOCA represent design base events necessary to verify the adequacy of the emergency core cooling system, the passive containment system and other safety systems. The validation of simulation results is therefore important to the IEC staff responsible for monitoring the research. (author); 3 refs

  20. Thermal-Hydraulic Integral Effect Test Result on the Steam Generator Tube Rupture Accident in the APR1400

    International Nuclear Information System (INIS)

    A postulated SGTR (Steam Generator Tube Rupture) event of the APR1400 was experimentally investigated with the thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation). It is generally known that the leak flow rate from the primary to the secondary side is the most important factor affecting the overall thermalhydraulic behaviors such as the depressurization rate of the RCS system, the water level increase and pressurization rate of the secondary system, and the consequent MSSV opening time, etc. As one of the most limiting SGTR accidents, a leak flow equivalent to a double-ended rupture of single and five U-tubes was simulated in this study. The main objectives of these tests were not only to provide physical insight into the system response of the APR1400 reactor during a transient situation of the SGTR but also to present integral effect test data for the validation of the SPACE (Safety and Performance Analysis Computer Code), which is now under development by the Korean nuclear industry

  1. Stability boundary calculation of thermal-hydraulic channels with RAMONA5, ATHLET and a reduced order model. A comparative study

    International Nuclear Information System (INIS)

    In the framework of the design study comprehensive system code analyses are performed with ATHLET and RAMONA. RAMONA is used in the current analysis because it has a broad validation basis for stability and transient analysis. On the other hand ATHLET has some advantages compared to RAMONA (free geometry and nodalization definition), which will be important to model and analyse the above mentioned test facility. One objective is to predict and confirm the operating conditions and transient behaviour for different facility designs. Thereby one aspect is the prediction of the thermal-hydraulic conditions at which self sustained density wave oscillations (ssDWOs) may occur under constant pressure drop boundary conditions. This paper is devoted to the latter investigation, only. In particular, we will discuss the question under which conditions the results of the measurement and simulation of the ssDWO onset are comparable to each other using the system codes ATHLET and RAMONA and, beside, an advanced reduced thermal-hydraulic model (TH-ROM). It will be shown why a precise measurement of the steady state axial profiles (void fraction, velocity of the liquid and gas phases and the axial pressure drop distribution) is of paramount importance in the scope of the present comparative study. (orig.)

  2. A coupled 3D neutron kinetics/thermal-hydraulics model of the generation IV sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    The Generation IV Sodium-cooled Fast Reactor (SFR) is an advanced fast-spectrum reactor concept being studied in the frame of international collaborations such as the Generation IV International Forum and European Union Framework Programmes. The present paper reports on the development and validation of a coupled 3D neutron kinetics / thermal-hydraulics model of a 3600 MWth SFR core being designed at CEA. The work has been performed in preparation for the analysis of transient core behavior in relation to hypothetical sodium boiling events, e.g. following an unprotected loss-of-flow (ULOF) accident or an unprotected transient overpower (UTOP) accident. The coupled 3D core model was developed in the frame of PSI's FAST code system, principally using the TRACE and PARCS codes. The neutronic data necessary for the 3D kinetics model in PARCS were derived from ERANOS-2.1 calculations. The standalone neutronics (PARCS) and thermal-hydraulics (TRACE) models were coupled by means of an external mapping scheme, and coupled simulations were performed to obtain steady-state and null-transient solutions for different core states. The principal neutronic parameters, mainly the effective multiplication factor and reactivity coefficients, were computed and validated against static ERANOS-2.1 calculations. Good agreement was obtained in each case. (authors)

  3. Stability boundary calculation of thermal-hydraulic channels with RAMONA5, ATHLET and a reduced order model. A comparative study

    Energy Technology Data Exchange (ETDEWEB)

    Jobst, Matthias; Lange, Carsten; Hennig, Dieter; Hurtado, Antonio [Technische Univ. Dresden (Germany). Professur fuer Wasserstoff- und Kernenergietechnik

    2013-07-01

    In the framework of the design study comprehensive system code analyses are performed with ATHLET and RAMONA. RAMONA is used in the current analysis because it has a broad validation basis for stability and transient analysis. On the other hand ATHLET has some advantages compared to RAMONA (free geometry and nodalization definition), which will be important to model and analyse the above mentioned test facility. One objective is to predict and confirm the operating conditions and transient behaviour for different facility designs. Thereby one aspect is the prediction of the thermal-hydraulic conditions at which self sustained density wave oscillations (ssDWOs) may occur under constant pressure drop boundary conditions. This paper is devoted to the latter investigation, only. In particular, we will discuss the question under which conditions the results of the measurement and simulation of the ssDWO onset are comparable to each other using the system codes ATHLET and RAMONA and, beside, an advanced reduced thermal-hydraulic model (TH-ROM). It will be shown why a precise measurement of the steady state axial profiles (void fraction, velocity of the liquid and gas phases and the axial pressure drop distribution) is of paramount importance in the scope of the present comparative study. (orig.)

  4. Thermal hydraulic and neutron kinetic simulation of the Angra 2 reactor using a RELAP5/PARCS coupled model

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Patricia A.L.; Costa, Antonella L.; Hamers, Adolfo R.; Pereira, Claubia; Rodrigues, Thiago D.A.; Mantecon, Javier G.; Veloso, Maria A.F., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: adolforomerohamers@hotmail.com, E-mail: claubia@nuclear.ufmg.br, E-mail: thiagodanielbh@gmail.com, E-mail: mantecon1987@gmail.com, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear

    2015-07-01

    The computational advances observed in the last two decades have been provided direct impact on the researches related to nuclear simulations, which use several types of computer codes, including coupled between them, allowing representing with very accuracy the behavior of nuclear plants. Studies of complex scenarios in nuclear reactors have been improved by the use of thermal-hydraulic (TH) and neutron kinetics (NK) coupled codes. This technique consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into codes, mainly to simulate transients that involve asymmetric core spatial power distributions and strong feedback effects between neutronics and reactor thermal-hydraulics. Therefore, this work presents preliminary results of TH RELAP5 and the NK PARCS calculations applied to model of the Angra 2 reactor. The WIMSD-5B code has been used to generate the macroscopic cross sections used in the NK code. The results obtained are satisfactory and represent important part of the development of this methodology. The next step is to couple the codes. (author)

  5. Thermal-hydraulic effects of transition to improved System 80TM fuel

    International Nuclear Information System (INIS)

    ABB CE's improved System 80TM PWR fuel design includes GUARDIAN debris-resistant features and laser-welded Zircaloy grids. The GUARDIAN features include an Inconel grid with debris-filtering features located just above the Lower End Fitting, and a solid fuel rod bottom end cap that extends above the filtering features. Tests and analyses were done to establish the impact of these design improvements on fuel assembly hydraulic performance. Further analysis was done to determine the mixed core thermal-hydraulic performance as the transition is made over two fuel cycles to a full core of the improved System 80TM fuel. Results confirm that the Thermal-Hydraulic (T-H) effects of the reduction in hydraulic resistance between the improved and resident fuel due to the laser-welded Zircaloy grids offsets the effects of the increased resistance GUARDIAN grid. Therefore, the mechanically improved System 80TM fuel can be implemented with no net impact on Departure from Nucleate Boiling (DNB) margin in transition cores. (author)

  6. Overview on CSNI Separate Effects Test Facility Matrices for Validation of Best Estimate Thermal-Hydraulic Computer Codes

    International Nuclear Information System (INIS)

    An internationally agreed separate effects test (Set) Validation Matrix for thermal-hydraulic system codes has been established by a sub-group of the Task Group on Thermal Hydraulic System Behaviour as requested by the OECD/Nea Committee on Safety of Nuclear Installations (CSNI) Principal Working Group No. 2 on Coolant System Behaviour. The construction of such a Matrix is an attempt to collect together in a systematic way the best sets of openly available test data for code validation, assessment and improvement and also for quantitative code assessment with respect to quantification of uncertainties to the modelling of individual phenomena by the codes. The methodology that has been developed during the process of establishing CSNI-Set validation matrix was an important outcome of the work on Set matrix. In the paper, the methodology developed will be discussed in detail, together with the examples from the Set matrix. In addition, all the choices, which have been made from the 187 identified facilities covering the 67 phenomena, will be investigated together with some discussions on the data-base. Facilities and phenomena have been cross-referenced in a separate effects test cross reference matrix, while the selected separate effects tests themselves are listed against the thermal-hydraulic phenomena for which they can provide validation data. As a preliminary to the classification of facilities and test data, it was necessary to identify a sufficiently complete list of relevant phenomena for LOCA and non-LOCA transient applications of PWRs and BWRs. The majority of these phenomena are also relevant to Advanced Water Cooled Reactors. To this end, 67 phenomena were identified for inclusion in the Set matrix and, in all; about 2094 tests are included in the Set matrix. The Set matrix, as it stands, is representative of the major part of the experimental work, which has been carried out in the LWR-safety thermal hydraulics field, covering a large number of

  7. Thermal hydraulics of rod bundles: The effect of eccentricity

    International Nuclear Information System (INIS)

    Highlights: • Present CFD investigation explores, whole bundle eccentricity for the first time. • Fluid flow and thermal characteristics in various subchannels are analyzed. • Mass flux distribution is particularly analyzed to study eccentricity effect. • Higher eccentricity resulted in a shoot up in rod surface temperature distribution. • Both tangential and radial flow in rod bundles has resulted due to eccentricity. -- Abstract: The effect of eccentricity on the fluid flow and heat transfer through a 19-rod bundle is numerically carried out. When the whole bundle shifts downwards with respect to the outer (pressure) tube, flow redistribution happens. This in turn is responsible for changes in mass flux, pressure and differential flow development in various subchannels. The heat flux imposed on the surface of the fuel rods and the mass flux through the subchannels determines the coolant outlet temperatures. The simulations are performed for a coolant flow Reynolds number of 4 × 105. For an eccentricity value of 0.7, the mass flux in the bottom most subchannel (l) was found to decrease by 10%, while the surface temperature of the fuel rod in the vicinity of this subchannel increased by 250% at the outlet section. Parameters of engineering interest including skin friction coefficient, Nusselt number, etc., have been systematically explored to study the effect of eccentricity on the rod bundle

  8. Thermal hydraulic study of a corium molten pool

    International Nuclear Information System (INIS)

    The thermohydraulic behaviour of a mass of molten core is investigated, in the frame of PWR severe accidents studies. The corium may be located in the vessel lower head or in an external core-catcher. It is assumed to be present in the container instantaneously. Its motion is described by one velocity field. It may be homogeneous or made of two stratified fluids. The residual power is assumed to be constant and uniform in the UO2 phase. The radiative losses and the external water-cooling are taken into account. The thermal resistance of a peripheral crust is considered. The influence of the crust on the pool geometry may be studied. The wall behaviour is analysed by a conduction calculation. The interest of a sacrificial layer is underlined, so as the necessity of a multicomponent multiphase model to study the behaviour of a core catcher. It is also concluded that some experiments are needed for code validation about volume heated natural convection and multiphase flows. (author). 14 figs., 3 refs

  9. Development of thermal hydraulic models for the reliable regulatory auditing code

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Song, C. H.; Lee, Y. J.; Kwon, T. S.; Lee, S. W. [Korea Automic Energy Research Institute, Taejon (Korea, Republic of)

    2004-02-15

    The objective of this project is to develop thermal hydraulic models for use in improving the reliability of the regulatory auditing codes. The current year fall under the second step of the 3 year project, and the main researches were focused on the development of downcorner boiling model. During the current year, the bubble stream model of downcorner has been developed and installed in he auditing code. The model sensitivity analysis has been performed for APR1400 LBLOCA scenario using the modified code. The preliminary calculation has been performed for the experimental test facility using FLUENT and MARS code. The facility for air bubble experiment has been installed. The thermal hydraulic phenomena for VHTR and super critical reactor have been identified for the future application and model development.

  10. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    1998-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is first step of the whole project, thus focus to the establishment of improvement area. The study was performed by reconsideration of the previous code assessment works and investigation of AECL design analysis tools. In order to identify the thermal hydraulic phenomena for events, the whole system of CANDU plant was divided into main functional systems and subcomponents. Each phenomena was addressed to the each subcomponent. Finally improvement areas of model development for auditing tool were established based on the identified phenomena. 8 refs., 21 figs., 19 tabs. (Author)

  11. Coupled thermal-hydraulic/aerosol transport analysis capability for severe accidents

    International Nuclear Information System (INIS)

    Fission product transport and thermal-hydraulic phenomena, occurring during the in-vessel phase of postulated accident progression, affect each other directly and significantly. It is important to couple the calculation of these processes to obtain accurate estimates of the magnitude and the timing of the release, of initial and revolatilized fission products from the primary system of a PWR, or the vessel of a BWR. The Electric Power Research Institute research efforts to obtain a functional and sophisticated coupled thermal-hydraulic/aerosol transport analysis capability are described. These are based on coupling the codes CORMLT, PSAAC (Primary System Accident Analysis Code) and RAFT (Reactor Aerosol Formation and Transport), which have been under development since 1982. Summary descriptions of these codes are also provided. (author)

  12. Large-scale Monte Carlo calculations with thermal-hydraulic feedback

    International Nuclear Information System (INIS)

    Monte Carlo based codes provide the most accurate solution of the particle transport problem. Individual particle trajectories are followed, and the interaction physics is simulated using detailed modeling of the physical reaction. The calculations are usually done using uniform temperature and density distributions. This is a significant approximation and leads to significantly distorted solution when applied to hot full power conditions. In this paper a method for introducing the thermal-hydraulic feedback by dynamic material distributions is introduced. The global variance reduction technique has been used to optimize the power tallying. The fission source convergence was accelerated by applying the Wielandt's acceleration method. Since the aim of this work is to solve coupled neutronic/thermal-hydraulic problems a convergence acceleration strategy based on stochastic approximation was proposed. The coupled system was applied to a quarter PWR core at pin and sub-channel level resolution. (author)

  13. Development and validation of thermal-hydraulic analysis code for plate type fuel research reactors

    International Nuclear Information System (INIS)

    A thermal-hydraulic analysis code has been developed with Visual Fortran 6.5 for the investigation of plate type fuel reactors, based on the fundamental conservation of mass, momentum and energy, and some proper constitutive correlations of flow friction factor, heat transfer and property. The Reactivity Insertion Accident(RIA) and Loss Of Flow Accident (LOFA), which have been defined in the IAEA 10 MW MTR Benchmark transients, were analyzed with this developed program. The comparison of some key parameters, such as the core power at scram, the maximum fuel temperature, the maximum clad temperature and the maximum outlet coolant temperature demonstrated that the results were consistent with that in the literature, which indicated that the model of this developed code is proper for thermal-hydraulic analysis of plate type research reactors. (authors)

  14. A generalized interface module for the coupling of spatial kinetics and thermal-hydraulics codes

    International Nuclear Information System (INIS)

    A generalized interface module has been developed for the coupling of any thermal-hydraulics code to any spatial kinetics code. The coupling scheme was designed and implemented with emphasis placed on maximizing flexibility while minimizing modifications to the respective codes. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine software to manage cross-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCX, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for two NEACRP rod ejection benchmark problems and an NEA/OECD main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated, nonetheless, the capabilities of the coupled code are presented for a typical PWR system/core model

  15. Current and anticipated uses of thermal-hydraulic codes in Germany

    International Nuclear Information System (INIS)

    In Germany, one third of the electrical power is generated by nuclear plants. ATHLET and S-RELAP5 are successfully applied for safety analyses of the existing PWR and BWR reactors and possible future reactors, e.g. EPR. Continuous development and assessment of thermal-hydraulic codes are necessary in order to meet present and future needs of licensing organizations, utilities, and vendors. Desired improvements include thermal-hydraulic models, multi-dimensional simulation, computational speed, interfaces to coupled codes, and code architecture. Real-time capability will be essential for application in full-scope simulators. Comprehensive code validation and quantification of uncertainties are prerequisites for future best-estimate analyses

  16. Condensation heat transfer coefficient with noncondensible gases for heat transfer in thermal hydraulic codes

    Energy Technology Data Exchange (ETDEWEB)

    Banerjee, S.; Hassan, Y.A. [Texas A& M Univ., College Station, TX (United States)

    1995-09-01

    Condensation in the presence of noncondensible gases plays an important role in the nuclear industry. The RELAP5/MOD3 thermal hydraulic code was used to study the ability of the code to predict this phenomenon. Two separate effects experiments were simulated using this code. These were the Massachusetts Institute of Technology`s (MIT) Pressurizer Experiment, the MIT Single Tube Experiment. A new iterative approach to calculate the interface temperature and the degraded heat transfer coefficient was developed and implemented in the RELAP5/MOD3 thermal hydraulic code. This model employs the heat transfer simultaneously. This model was found to perform much better than the reduction factor approach. The calculations using the new model were found to be in much better agreement with the experimental values.

  17. Thermal-hydraulic design concept of the solid-target system of spallation neutron source

    International Nuclear Information System (INIS)

    In relation to thermal-hydraulic design of the N-Arena solid-target system of the JHF project, heat transfer experiments were performed to obtain experimental data systematically on heat transfer coefficient and CHF for vertical upward and horizontal flows in a thin rectangular channel simulating a coolant channel of the proposed spallation neutron source. Thermal-hydraulic correlations which can be used for design calculations were proposed based on the obtained data. Finally tentative results of feasibility study on maximum beam power which could be attained with a solid target were presented. The result indicated that the condition for the onset of nucleate boiling is the most significant limiting factor to the maximum beam power. (author)

  18. A parametric thermal-hydraulic analysis of I.T.U. TRIGA Mark-II reactor

    International Nuclear Information System (INIS)

    In this study, a transient, one-dimensional thermal-hydraulic subchannel analysis for I.T.U. TRIGA Mark-II reactor was employed. The cooling of this reactor is based on natural convection; however, mixed convection is considered in modeling in order to enhance the capability of the computer code. After the continuity, conservation of energy, momentum balance equations for coolant in axial direction and heat conduction equation for fuel rod in radial direction had been written, they were discretized by using the control volume approach to obtain a set of algebraic equations. By the aid of discretized continuity and momentum balance equations, a pressure correction equation was derived. Then, a FORTRAN program called TRIGATH (TRIGA Thermal-Hydraulics) has been developed to solve this set of algebraic equations by using SIMPLE algorithm. As a result, the temperature distributions of the coolant and fuel rods as well as the velocity and pressure distributions of the coolant have been estimated. (authors)

  19. Sodium natural convection testing in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility

    International Nuclear Information System (INIS)

    A comparison is made between experimental data and analytical results for a single-phase natural convection test in an experimental sodium loop. The test was conducted in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility, an engineering-scale high temperature sodium loop at the Oak Ridge National Laboratory (ORNL), used for thermal-hydraulic testing of simulated Liquid Metal Fast Breeder Reactor (LMFBR) subassemblies at normal and off-normal operating conditions. Electrical heating in the 19-pin assembly during the test was typical of decay heat levels. The test chosen for analysis in this paper was one of seven natural convection runs conducted in the facility. In this test the bypass line was open to simulate a parallel heated assembly and the test was begun with a pump coastdown from a small initial forced flow

  20. Condensation heat transfer coefficient with noncondensible gases for heat transfer in thermal hydraulic codes

    International Nuclear Information System (INIS)

    Condensation in the presence of noncondensible gases plays an important role in the nuclear industry. The RELAP5/MOD3 thermal hydraulic code was used to study the ability of the code to predict this phenomenon. Two separate effects experiments were simulated using this code. These were the Massachusetts Institute of Technology's (MIT) Pressurizer Experiment, the MIT Single Tube Experiment. A new iterative approach to calculate the interface temperature and the degraded heat transfer coefficient was developed and implemented in the RELAP5/MOD3 thermal hydraulic code. This model employs the heat transfer simultaneously. This model was found to perform much better than the reduction factor approach. The calculations using the new model were found to be in much better agreement with the experimental values

  1. Thermal hydraulics characterization of the core and the reactor vessel type BWR

    International Nuclear Information System (INIS)

    The thermal hydraulics design of a reactor type BWR 5 as the employees in the nuclear power plant of Laguna Verde involves the coupling of at least six control volumes: Pumps jet region, Stratification region, Core region, Vapor dryer region, Humidity separator region and Reactor region. Except by the regions of the core and reactor, these control volumes only are used for design considerations and their importance as operative data source is limited. It is for that is fundamental to complement the thermal hydraulics relations to obtain major data that allow to determine the efficiency of internal components, such as pumps jet, humidity separator and vapor dryer. Like example of the previous thing, calculations are realized on the humidity of the principal vapor during starting, comparing it with the values at the moment incorporated in the data banks of the computers of process of both units. (Author)

  2. Thermal hydraulic calculations of the IRT-Sofia reactor with LEU fuel IRT-4M, Sofia

    International Nuclear Information System (INIS)

    The neutronic calculations of the IRT-Sofia research reactor with fuel assemblies (FA) of the IRT-4M type, containing low enriched uranium (19.75%), confirmed the justness of the selection of its initial core configuration. On the basis of the neutronic calculation results the thermal-hydraulic calculations were done by PLTEMP code. Three possible operational regimes have been considered. For each of them the margin coefficients of the water onset of the nucleate boiling (ONB) have been determined. An analysis has been made on the fuel element surface for the particular fuel assembly with maximum power density. The calculations have been accomplished for the core water inlet temperature of 450C. The results are in compliance with the thermal-hydraulic safety requirements. This is valid for all analyzed regimes

  3. Thermal-hydraulic analysis of tajoura research reactor during primary loop pumps failure

    International Nuclear Information System (INIS)

    One of the main requirements of the safety analysis report of tajoura research reactor is the thermal hydraulic analysis and the determination of the maximum fuel surface (clad) temperatures during the cut-off of the electrical power and failure of the primary circuit pumps. This paper is concerned with the thermal hydraulic analysis of the tajoura reactor and the calculation of the fuel and coolant temperatures. The objective of the study is to asses the possibility of cooling the reactor core by the emergency tank during the 1st minute after pumps failure and by natural convection after the automatic opening of the natural convection valves. the results of the present study show that the reactor is cooled properly by the primary circuit flow during normal operation with the maximum power of 10 MW, and that the emergency tank and natural convection provide sufficient cooling of the reactor after pump failure

  4. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-04-15

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is first step of the whole project, thus focus to the establishment of improvement area. The study was performed by reconsideration of the previous code assessment works and investigation of AECL design analysis tools. In order to identify the thermal hydraulic phenomena for events, the whole system of CANDU plant was divided into main functional systems and subcomponents. Each phenomena was addressed to the each subcomponent. FinaIly improvement areas of model development for auditing tool were established based on the identified phenomena.

  5. Current and anticipated uses of thermal-hydraulic codes in Germany

    Energy Technology Data Exchange (ETDEWEB)

    Teschendorff, V.; Sommer, F.; Depisch, F.

    1997-07-01

    In Germany, one third of the electrical power is generated by nuclear plants. ATHLET and S-RELAP5 are successfully applied for safety analyses of the existing PWR and BWR reactors and possible future reactors, e.g. EPR. Continuous development and assessment of thermal-hydraulic codes are necessary in order to meet present and future needs of licensing organizations, utilities, and vendors. Desired improvements include thermal-hydraulic models, multi-dimensional simulation, computational speed, interfaces to coupled codes, and code architecture. Real-time capability will be essential for application in full-scope simulators. Comprehensive code validation and quantification of uncertainties are prerequisites for future best-estimate analyses.

  6. An approach to validation of coupled CFD and system thermal-hydraulics codes

    International Nuclear Information System (INIS)

    This paper discusses the development of approach and experimental facility for the validation of coupled Computational Fluid Dynamics (CFD) and System Thermal Hydraulics (STH) codes. The validation of a coupled code requires experiments which feature two way feedback between the component (CFD sub-domain) and the system (STH sub-domain). We present results of CFD analysis that are used in the development of a flexible design for the TALL-3D experimental facility. The facility consists of a lead-bismuth thermal-hydraulic loop operating in forced and natural circulation regimes with a heated pool-type 3D test section. The goal of the design is to achieve a feedback between mixing and stratification phenomena in the 3D tests section and forced / natural circulation flow conditions in the loop. Finally, we discuss the development of an experimental validation matrix for validation of coupled STH and CFD codes that considers the key physical phenomena of interest. (author)

  7. Thermal hydraulics of sodium-cooled fast reactors - key issues and highlights

    International Nuclear Information System (INIS)

    In this paper key issues and highlighted topics in thermal hydraulics are discussed in connection to the current Japan's sodium-cooled fast reactor development efforts. In particular, design study and related researches of the Japan Sodium-cooled Fast Reactor (JSFR) are focused. Several innovative technologies, e.g., compact reactor vessel, two-loop system, fully natural circulation decay heat removal, and recriticality free core, have been investigated in order to reduce construction cost and to achieve higher level of reactor safety. Preliminary evaluations of innovative technologies to be applied to JSFR are on-going. Here, progress of design study is introduced. Then, research and development activities on the thermal hydraulics related to the innovative technologies are briefly reviewed. (author)

  8. Assessment of multi-scale thermal-hydraulic simulation for a PWR steam generator with CUPID/MARS coupling

    International Nuclear Information System (INIS)

    For the thermal hydraulics analysis of nuclear reactor components of LWRs (Light Water Reactors) such as reactor vessel, steam generator, containment, a multi-dimensional two-phase flow code, named CUPID, has been being developed. The CUPID code pursues a capability of multi-physics and multi-scale thermal hydraulics analysis. In the present study, multi-scale simulation was performed by coupling with system-scale code, MARS. The coupled code was assessed to visualize the flow behavior of the steam generator of the Advanced Power Reactor (APR1400). The primary side of the steam generator and remaining Reactor Coolant System (RCS) is modeled by MARS and secondary side is by CUPID. For the secondary side simulation by the CUPID part, a porous media approach was adopted to two-fluid model and conductor model to simplify the complicated geometry of the steam generator. In order to obtain a porosity of a given computing cell, a special algorithm was employed to directly calculate volume ratio by mapping the 3D CAD file onto the grid system. Besides, the proper constitutive relationships for U-tubes are considered further. To treat the complex thermo-hydraulic phenomena on the shell side of a steam generator, a set of constitutive models available in the literature for a two-phase flow map, interfacial heat and mass transfer, interfacial drag, wall friction, wall heating, and heat partitioning in flows over tube bundles were applied to close the numerical model. This paper presents the description of the coupling method, porous media approach to simplify the steam generator, and the simulation results using the coupled codes. (author)

  9. RESAQ - a thermal hydraulics program of warming up and cooling down for PWR type reactors

    International Nuclear Information System (INIS)

    RESAQ is a computer code that is being developed for the thermal hydraulic analysis of the warming up and cooling down procedures for PWR nuclear reactors. The mass and energy balances applied to control volumes yield a 42 equations system that is solved using Newton-Raphson method and Gauss elimination. Several restrictions applied to the process as maximum temperature variation at the surgeline, maximum rate of pressurization or depressurization are considered in the code. (author)

  10. Reactor thermal-hydraulic FY 1986 status report for the multimegawatt Space Nuclear Power Program

    International Nuclear Information System (INIS)

    PNL's 1986 activities can be divided into three basic areas: code assessment, correlation assessment and experimental activities. The ultimate goal of all these activities is developing computer codes and verifying their use to perform the thermal-hydraulic analysis and design of the reactor core and plenum of the various proposed concepts. To perform this task as assessment is made of existing computer codes, models, correlations, and microgravity experimental data

  11. Overview of MAAP thermal-hydraulic qualifications and plant application guidelines

    International Nuclear Information System (INIS)

    the Modular Accident Analysis Program (MAAP) was developed under IDCOR (Industry Degraded Core Rule-Making) sponsorship to simulate light water reactor (LWR) system response to accident initiating events. Electric Power Research institute (EPRI) has continued its development since 1985, and at the conclusion of the IDCOR Program, EPRI has assumed full responsibility for MAAP. In addition to the evaluation of the outcome of postulated severe accidents in LWRs, MAAP has the versatility to extend its application to analyze effects of operator actions during the progression of an accident to assess accident management procedures. Since the thermal-hydraulic phenomena govern the system response during the early phase of the transient a need exists to quantify thermal-hydraulic models and associated application of MAAP for industry use. Therefore, the objective of the project is to qualify and demonstrate the thermal-hydraulic modeling capabilities of MAAP for its anticipated applications by utility engineers. The industry has supported an extensive research and development effort that generated a database from small-scale to large-scale tests for phenomenological understanding of thermal-hydraulic behavior of LWRs. Such efforts include simulation of transient response for both pressurized water reactor (PWR) and boiling water reactor (BWR) plants. In addition EPRI has supported work in acquiring a plant data base. This work will focus on using the existing data and available information for MAAP validation. A frozen version of MAAP code (MAAP 3.0B) for PWR and BWR is selected for validation. The project tasks are described

  12. The Penn State Nodal Expansion Transient Analysis Technique with thermal-hydraulic feedback

    International Nuclear Information System (INIS)

    The nuclear engineering department of the Pennsylvania State University has under development a nodal neutron kinetics code. The PEnn State Nodal Expansion TRansient Analysis TEchnique (PENETRATE) performs two-group, three-dimensional nodal kinetics calculations using the nodal expansion method (NEM). The focus of this discussion is its performance in the solution of the Langenbuch-Maurer-Werner light water rector (LMW LWR) problem. This transient requires an accurate model of both control rod motion and coupled thermal-hydraulic feedback

  13. ITHNA.SYS: An Integrated Thermal Hydraulic and Neutronic Analyzer SYStem for NUR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mazidi, S., E-mail: samirmazidi@gmail.com [Division Physique et Applications Nucléaires, Centre de Recherche Nucléaire de Draria (CRND), BP 43 Sebala, Draria, Alger (Algeria); Meftah, B., E-mail: b_meftah@yahoo.com [Division Physique et Applications Nucléaires, Centre de Recherche Nucléaire de Draria (CRND), BP 43 Sebala, Draria, Alger (Algeria); Belgaid, M., E-mail: belgaidm@yahoo.com [Faculté de Physique, Université Houari Boumediene, USTHB, BP 31, Bab Ezzouar, Alger (Algeria); Letaim, F., E-mail: fletaim@yahoo.fr [Faculté des Sciences et Technologies, Université d’El-oued, PO Box 789, El-oued (Algeria); Halilou, A., E-mail: hal_rane@yahoo.fr [Division Réacteur NUR, Centre de Recherche Nucléaire de Draria, BP 43 Sebala, Draria, Alger (Algeria)

    2015-08-15

    Highlights: • We develop a neutronic and thermal hydraulic MTR reactor analyzer. • The analyzer allows a rapid determination of the reactor core parameters. • Some NUR reactor parameters have been analyzed. - Abstract: This paper introduces the Integrated Thermal Hydraulic and Neutronic Analyzer SYStem (ITHNA.SYS) that has been developed for the Algerian research reactor NUR. It is used both as an operating aid tool and as a core physics engineering analysis tool. The system embeds three modules of the MTR-PC software package developed by INVAP SE: the cell calculation code WIMSD, the core calculation code CITVAP and the program TERMIC for thermal hydraulic analysis of a material testing reactor (MTR) core in forced convection. ITHNA.SYS operates both in on-line and off-line modes. In the on-line mode, the system is linked, via the computer parallel port, to the data acquisition console of the reactor control room and allows a real time monitoring of major physical and safety parameters of the NUR core. PC-based ITHNA.SYS provides a viable and convenient way of using an accumulated and often complex reactor physics stock of knowledge and frees the user from the intricacy of adequate reactor core modeling. This guaranties an accurate, though rapid, determination of a variety of neutronic and thermal hydraulic parameters of importance for the operation and safety analysis of the NUR research reactor. Instead of the several hours usually required, the processing time for the determination of such parameters is now reduced to few seconds. Validation of the system was performed with respect to experimental measurements and to calculations using reference codes. ITHNA.SYS can be easily adapted to accommodate other kinds of MTR reactors.

  14. PLUGM: a coupled thermal-hydraulic computer model for freezing melt flow in a channel

    International Nuclear Information System (INIS)

    PLUGM is a coupled thermal-hydraulic computer model for freezing liquid flow and plugging in a cold channel. PLUGM is being developed at Sandia National Laboratories for applications in Sandia's ex-vessel Core Retention Concept Assessment Program and in Sandia's LMFBR Transition Phase Program. The purpose of this paper is to introduce PLUGM and demonstrate how it can be used in the analysis of two of the core retention concepts under investigation at Sandia: refractory brick crucibles and particle beds

  15. Computer code for the thermal-hydraulic analysis of ITU TRIGA Mark-II reactor

    International Nuclear Information System (INIS)

    Istanbul Technical University (ITU) TRIGA Mark-II reactor core consists of ninety vertical cylindrical elements located in five rings. Sixty-nine of them are fuel elements. The reactor is operated and cooled with natural convection by pool water, which is also cooled and purified in external coolant circuits by forced convection. This characteristic leads to consider both the natural and forced convection heat transfer in a 'porous-medium analysis'. The safety analysis of the reactor requires a thermal-hydraulic model of the reactor to determine the thermal-hydraulic parameters in each mode of operation. In this study, a computer code cooled TRIGA-PM (TRIGA - Porous Medium) for the thermal-hydraulic analysis of ITU is considered. TRIGA Mark-II reactor code has been developed to obtain velocity, pressure and temperature distributions in the reactor pool as a function of core design parameters and pool configuration. The code is a transient, thermal-hydraulic code and requires geometric and physical modelling parameters. In the model, although the reactor is considered as only porous medium, the other part of the reactor pool is considered partly as continuum and partly as porous medium. COMMIX-1C code is used for the benchmark purpose of TRIGA-PM code. For the normal operating conditions of the reactor, estimations of TRIGA-PM are in good agreement with those of COMMIX-1C. After some more improvements, this code will be employed for the estimation of LOCA scenario, which can not be analyses by COMMIX-1C and the other multi-purpose codes, considering a break at one of the beam tubes of the reactor

  16. Large-scale Monte Carlo neutron transport calculations with thermal hydraulic feedback

    International Nuclear Information System (INIS)

    Highlights: • Method of internal coupling, based on dynamic material distribution, is presented. • The Wielandt shift method is implemented to accelerate Mote Carlo calculations. • The Uniform Fission Site method is introduced for tallies with large numbers of bins. • The stochastic approximation scheme is used to stabilize coupled code convergence. - Abstract: The Monte Carlo method provides the most accurate description of the particle transport problem. The criticality problem is simulated by following the histories of individual particles without approximating the energy, angle or the coordinate dependence. These calculations are usually done using homogeneous thermal hydraulic conditions. This is a very crude approximation in the general case. In this paper, the method of internal coupling between neutron transport and thermal hydraulics is presented. The method is based on dynamic material distribution, where coordinate dependent temperature and density information is supplied on the fly during the transport calculation. This method does not suffer from the deficiencies characteristic of the external coupling via the input files. In latter case, the geometry is split into multiple cells having distinct temperatures and densities to supply the feedback. The possibility to efficiently simulate large scale geometries at pin-by-pin and subchannel level resolution was investigated. The Wielandt shift method for reducing the dominance ratio of the system and accelerating the fission source convergence was implemented. During the coupled iteration a detailed distribution of the fission heat deposition is required by the thermal hydraulics calculation. Providing reasonable statistical uncertainties for tallies having large numbers of bins, is a complicated task. This problem was resolved by applying the Uniform Fission Site method. Previous investigations showed that the convergence of the coupled neutron transport/thermal hydraulics calculation is limited by

  17. Interface between thermal-hydraulics and fission product transport in severe accident analysis

    International Nuclear Information System (INIS)

    Studies of the effectiveness of the containment following hypothetical severe accidents in LWRs have traditionally been split into two separate disciplines. The first is the thermal-hydraulics of the containment atmosphere, leading to an estimate of the magnitudes of the threats to the integrity of the containment building. The second is the study of the transport of the fission products released from the core within and from the building, the ultimate output of which, given a building damage state, is a radiological source term to the environment. In fact, the transport of fission products is strongly influenced by the prevailing thermal-hydraulic conditions, and there are some ways in which the fission products have an influence back on the thermal-hydraulics. The paper describes work funded and co-ordinated by the Commission of the European Communities (CEC) to investigate how the two sorts of calculation, the thermal-hydraulic and the fission product transport, can be made more responsive to the needs of each other. Recent theoretical and experimental work has indicated that the deposition of fission product aerosols is markedly enhanced by the condensation of steam on the particles. The rate of such condensation is a sensitive function of the thermodynamic conditions in the atmosphere, and the release of latent heat because of condensation can in turn significantly change these conditions. The rates of condensation are increased if the particles contain chemical components which are water soluble. To model this effect one requires considerable knowledge of the thermodynamic properties of the relevant aqueous solutions. Recent experience has shown that considerable differences occur between the predictions of different codes. In order to resolve these discrepancies, the CEC has organised a code comparison exercise based on the LACE LA-4 experiment and is currently organising another code comparison based on one of the experiments in the DEMONA series. (author

  18. Test program element II blanket and shield thermal-hydraulic and thermomechanical testing, experimental facility survey

    Energy Technology Data Exchange (ETDEWEB)

    Ware, A.G.; Longhurst, G.R.

    1981-12-01

    This report presents results of a survey conducted by EG and G Idaho to determine facilities available to conduct thermal-hydraulic and thermomechanical testing for the Department of Energy Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program. In response to EG and G queries, twelve organizations (in addition to EG and G and General Atomic) expressed interest in providing experimental facilities. A variety of methods of supplying heat is available.

  19. ITHNA.SYS: An Integrated Thermal Hydraulic and Neutronic Analyzer SYStem for NUR research reactor

    International Nuclear Information System (INIS)

    Highlights: • We develop a neutronic and thermal hydraulic MTR reactor analyzer. • The analyzer allows a rapid determination of the reactor core parameters. • Some NUR reactor parameters have been analyzed. - Abstract: This paper introduces the Integrated Thermal Hydraulic and Neutronic Analyzer SYStem (ITHNA.SYS) that has been developed for the Algerian research reactor NUR. It is used both as an operating aid tool and as a core physics engineering analysis tool. The system embeds three modules of the MTR-PC software package developed by INVAP SE: the cell calculation code WIMSD, the core calculation code CITVAP and the program TERMIC for thermal hydraulic analysis of a material testing reactor (MTR) core in forced convection. ITHNA.SYS operates both in on-line and off-line modes. In the on-line mode, the system is linked, via the computer parallel port, to the data acquisition console of the reactor control room and allows a real time monitoring of major physical and safety parameters of the NUR core. PC-based ITHNA.SYS provides a viable and convenient way of using an accumulated and often complex reactor physics stock of knowledge and frees the user from the intricacy of adequate reactor core modeling. This guaranties an accurate, though rapid, determination of a variety of neutronic and thermal hydraulic parameters of importance for the operation and safety analysis of the NUR research reactor. Instead of the several hours usually required, the processing time for the determination of such parameters is now reduced to few seconds. Validation of the system was performed with respect to experimental measurements and to calculations using reference codes. ITHNA.SYS can be easily adapted to accommodate other kinds of MTR reactors

  20. Influence of Thermal-hydraulic Model to Fuel Management Core Calculation

    International Nuclear Information System (INIS)

    The integration of neutronic, fuel rod and thermal-hydraulic calculations for both, steady-state core design type of the calculation and for transient and safety analyses, is used to improve the response of Nuclear Power Plants (NPP) both from the point of view of safe and economic plant operation. That process assumes improvement of current calculational tools and application of results acquired from operational experience. The objective of this paper is to explore influence of improved thermal-hydraulics core model to overall in-core fuel management parameters (reactivity, power distribution, burn-up distribution) and to take it into account in systematic way. New core thermal-hydraulics model based on codes COBRA III C and COBRA-EN was included within the PARCS depletion loop to calculate the behavior of representative fuel rods for each assembly. Modified code is called COBRA-VIP and exists as both standalone version and part of PARCS code. Some programming changes were necessary to make possible dual use of COBRA-VIP. Core fuel management calculation was performed for NPP Krsko cycle 23 to show influence of performed change to selected core parameters. The benefit of the described approach is that in addition to normal depletion calculation, the behavior of any fuel rod in each fuel assembly can be studied from thermal-hydraulics point of view. The average fuel rod per assembly can be used to improve TH feedback calculations and the limiting fuel rod per assembly can be used to perform DNBR or fuel centre line temperature calculation.(author)

  1. Test program element II blanket and shield thermal-hydraulic and thermomechanical testing, experimental facility survey

    International Nuclear Information System (INIS)

    This report presents results of a survey conducted by EG and G Idaho to determine facilities available to conduct thermal-hydraulic and thermomechanical testing for the Department of Energy Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program. In response to EG and G queries, twelve organizations (in addition to EG and G and General Atomic) expressed interest in providing experimental facilities. A variety of methods of supplying heat is available

  2. A study of thermal stratification in the cold legs during the subcooled blowdown phase of a loss of coolant accident in the OSU APEX thermal hydraulic testing facility

    International Nuclear Information System (INIS)

    Thermal stratification, which has been linked to the occurrence of pressurized thermal shock (PTS), is observed to occur during the early stages of simulated loss of coolant accidents (LOCAS) in the Oregon State University Advanced Plant Experiment (OSU APEX) Thermal Hydraulic Test Facility. The OSU APEX Test Facility is a scaled model of the Westinghouse AP600 nuclear power plant. Analysis of the OSU APEX facility data has allowed the determination of an onset criteria for thermal stratification and has provided support for the postulated mechanisms leading to thermal stratification. CFX 4.1, a computational fluid dynamics code, was used to generate a model of the cold legs and the downcomer that described the phenomena occurring within them. Some mixing phenomena were predicted that lead to non-uniformity between the two cold legs attached to the steam generator on the side of the facility containing the Passive Residual Heat Removal (PRHR) injection system. The stratification was found to be two phase and unlikely to be a factor in PTS

  3. MATRA-LMR code for thermal-hydraulic subchannel analysis of LMR

    International Nuclear Information System (INIS)

    In the sodium cooled liquid metal reactors, the core thermal-hydraulic design limits are imposed on the maximum temperatures of claddings and fuel pins. Thus an accurate prediction of core coolant/fuel temperature distributions is essential to the LMR core thermal-hydraulic design. The detailed subchannel thermal-hydraulic analysis code MATRA-LMR is developed for KALIMER core subassembly design and analysis based on COBRA-IV-I and MATRA. To assess the developement status of this code, benchmark calculations have been performed for ORNL 19-pin assembly test and EBR-II 91-pin experiment. The bechmark analysis involved comparisons to the measured values and the calculated results. The calculation results of MATRA-LMR were compared to the measurements, and to SABRE4 and SLTHEN code calculation results, respectively. Finally, the applicability and practicality of the code have been shown by comparing the temperature calculation results of MATRA-LMR, SABRE4 and SLTHEN codes for a representative KALIMER core subassemblies

  4. Thermal hydraulic performance assessment of 18x18 solid fuel for the reactor power uprate

    Energy Technology Data Exchange (ETDEWEB)

    Shin, C. H.; Kim, H. I.; In, W. K.; Chun, T. H. [KAERI, Daejeon (Korea, Republic of)

    2009-07-01

    The thermal hydraulic analysis of the 18x18 solid fuel assembly has been carried out for the power uprate of OPR-1000. The suggested 18x18 solid fuel assembly has a structural compatibility for reloading to operating PWR reactors of OPR-1000. In the thermal hydraulic analysis, the mixing effect between the neighboring channels has to be carried out in a subchannel analysis. A subchannel analysis code, MATRA has been developed by KAERI. The main parameters for the thermal hydraulic design, such as a pressure drop and DNBR, and the maximum temperature in a fuel pellet have been estimated. The 18x18 solid fuel in the 120% power uprate showed an increased pressure drop and a similar DNBR behavior. The peak temperature in the fuel centerline, however, was slightly higher than that of the 16x16 solid fuel assembly for the normal operation condition. The peak temperature of a fuel pellet in the solid fuel should be seriously considered for increasing power density.

  5. Thermal-hydraulic and neutronic considerations for designing a lithium-cooled tokamak blanket

    International Nuclear Information System (INIS)

    A methodology for the design of lithium cooled blankets is developed. The thermal-hydraulics, neutronics and interactions between them are extensively investigated. In thermal hydraulics, two models illustrate the methodology used to obtain the acceptable ranges for a set of design parameters. The methodology can be used to identify the limiting constraints for a particular design. A complete neutronic scheme is set up for the calculations of the volumetric heating rate as a function of the distance from the first wall, the breeding ratio as a function of the amount of structural material in the blanket, and the radiation damage in terms of atom displacements and gas production rate. Different values of the volume percent of Type-316 stainless steel are assigned in four breeding zones to represent a nonuniformly distributed structural material which satisfies various thermal-hydraulic requirements. The role that the radiation damage plays in the overall design methodology is described. The product of the first wall lifetime and neutron loading is limited by the radiation damage which degrades the mechanical properties of the material

  6. CFX analysis of the CANDU moderator thermal-hydraulics in the Stern Lab. Test Facility

    International Nuclear Information System (INIS)

    A numerical calculation with the commercial CFD code CFX is conducted for a test facility simulating the CANDU moderator thermal-hydraulics. Two kinds of moderator thermal-hydraulic tests at Stern Laboratories Inc. were performed in the full geometric configuration of the CANDU moderator circulating vessel, which is called a Calandria, housing a matrix of horizontal rod bundles simulating the Calandria tubes. The first of these tests is the pressure drop measurement of a cross flow in the horizontal rod bundles. The other is the local temperature measurement on the cross section of the horizontal cylinder vessel simulating the Calandria. In the present study the full geometric details of the Calandria are incorporated in the grid generation of the computational domain to which the boundary conditions for each experiment are applied. The numerical solutions are reviewed and compared with the available test data. It is shown that the present CFD prediction without the empirical correlation based on the pressure drop test is in good agreement with the test results. The prediction becomes more accurate, as the flow conditions become more turbulent with a higher Reynolds number. However, the temperature fluctuation is observed during iteration steps for a steady-state simulation of the thermal-hydraulic test. This result shows that the flow and temperature distribution inside the moderator tank may not be stable in the actual test

  7. Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, J.; Park, W. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

  8. Evaluation on thermal-hydraulic characteristics for passive safety device of APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Seong Yeon; Lee, S. H.; Son, M. K. [Korea Association for Nuclear Technology, Taejon (Korea, Republic of); Jee, M. S.; Chung, M. H. [Chungnam National Univ., Taejon (Korea, Republic of)

    2001-07-15

    To establish evaluation and verification guideline for the APR1400, thermal-hydraulic characteristics for fuel rod bundle, reactor vessel and fluidic device is analyzed using FLUENT. Scope and major results of research are as follows : Thermal-hydraulic characteristics for nuclear fuel rod bundle: design data for nuclear fuel rod bundle and structure are surveyed, and 3 x 3 sub-channel model is adopted to investigate the fluid flow and heat transfer characteristics in fuel rod bundle. Computational results are compared with the heat transfer data measured by naphthalene sublimation method, and numerical analysis and evaluation are performed at various design conditions and flow conditions. Thermal-hydraulic characteristics for reactor vessel: reactor vessel design data are surveyed to develop numerical model. Porous media model is applied for fuel rod bundle, and full-scale, three dimensional simulation is performed at actual operating conditions. Distributions of velocity, pressure and temperature are discussed. Flow characteristics for fluidic device: three dimensional numerical model for fluidic device is developed, and numerical results are compared with experimental data obtained at KAERI in order to verify numerical simulation. In addition, variation of flow rate is investigated at various elapsed times after valve operating, and flow characteristics is analyzed at low and high flow rate conditions, respectively.

  9. Thermal hydraulic analysis of 3 MW TRIGA research reactor of bangladesh considering different cycles of burnup

    International Nuclear Information System (INIS)

    Burnup dependent steady state thermal hydraulic analysis of TRIGA Mark-II research reactor has been carried out utilizing coupled point kinetics, neutronics and thermal hydraulics code EUREKA-2/RR. From the previous calculations of neutronics parameters including percentage burnup of individual fuel elements performed so far for 700 MWD burnt core of TRIGA reactor showed that the fuel rod predicted as hottest at the beginning of cycle (fresh core) was found to remain as the hottest until 200 MWD of burn, but, with the progress of core burn, the hottest rod was found to be shifted and another rod in the core became the hottest. The present study intends to evaluate the thermal hydraulic parameters of these hottest fuel rods at different cycles of burnup, from beginning to 700 MWD core burnt considering reactor operates under steady state condition. Peak fuel centerline temperature, maximum cladding and coolant temperatures of the hottest channels were calculated. It revealed that maximum temperature reported for fuel clad and fuel centerline found to lie below their melting points which indicate that there is no chance of burnout on the fuel cladding surface and no blister in the fuel meat throughout the considered cycles of core burnt. (author)

  10. Analytical benchmarks for verification of thermal-hydraulic codes based on sub-channel approach

    International Nuclear Information System (INIS)

    Over the last year (2007), preliminary tests have been performed on the Moroccan TRIGA MARK II research reactor to show that, under all operating conditions, the coolant parameters fall within the ranges allowing the safe working conditions of the reactor core. In parallel, a sub-channel thermal-hydraulic code, named SACATRI (Sub-channel Analysis Code for Application to TRIGA reactors), was developed to satisfy the needs of numerical simulation tools, able to predict the coolant flow parameters. The thermal-hydraulic model of SACATRI code is based on four partial differential equations that describe the conservation of mass, energy, axial and transversal momentum. However, to achieve the full task of any numerical code, verification is a highly recommended activity for assessing the accuracy of computational simulations. This paper presents a new procedure which can be used during code and solution verification activities of thermal-hydraulic tools based on sub-channel approach. The technique of verification proposed is based mainly on the combination of the method of manufactured solution and the order of accuracy test. The verification of SACATRI code allowed the elaboration of exact analytical benchmarks that can be used to assess the mathematical correctness of the numerical solution to the elaborated model

  11. Analytical benchmarks for verification of thermal-hydraulic codes based on sub-channel approach

    Energy Technology Data Exchange (ETDEWEB)

    Merroun, O. [LMR/ERSN, Department of Physics, Faculty of Sciences, Abdelmalek Essaadi University, B.P. 2121, Tetouan 93002 (Morocco)], E-mail: meroun.ossama@gmail.com; Almers, A. [Department of Energetics, Ecole Nationale Superieure d' Arts et Metiers, Moulay Ismail University, B.P. 4024, Meknes (Morocco); El Bardouni, T.; El Bakkari, B. [LMR/ERSN, Department of Physics, Faculty of Sciences, Abdelmalek Essaadi University, B.P. 2121, Tetouan 93002 (Morocco); Chakir, E. [LRM/EPTN, Department of Physics, Faculty of Sciences, Kenitra (Morocco)

    2009-04-15

    Over the last year (2007), preliminary tests have been performed on the Moroccan TRIGA MARK II research reactor to show that, under all operating conditions, the coolant parameters fall within the ranges allowing the safe working conditions of the reactor core. In parallel, a sub-channel thermal-hydraulic code, named SACATRI (Sub-channel Analysis Code for Application to TRIGA reactors), was developed to satisfy the needs of numerical simulation tools, able to predict the coolant flow parameters. The thermal-hydraulic model of SACATRI code is based on four partial differential equations that describe the conservation of mass, energy, axial and transversal momentum. However, to achieve the full task of any numerical code, verification is a highly recommended activity for assessing the accuracy of computational simulations. This paper presents a new procedure which can be used during code and solution verification activities of thermal-hydraulic tools based on sub-channel approach. The technique of verification proposed is based mainly on the combination of the method of manufactured solution and the order of accuracy test. The verification of SACATRI code allowed the elaboration of exact analytical benchmarks that can be used to assess the mathematical correctness of the numerical solution to the elaborated model.

  12. Modeling of thermal hydraulics behaviour in reactor core of reactor TRIGA PUSPATI (RTP)

    International Nuclear Information System (INIS)

    Reactor TRIGA PUSPATI (RTP) in Malaysian Nuclear Agency (Nuclear Malaysia) is the one and only research reactor in Malaysia and had been used exclusively for research and development (R and D), training for reactor operators and education purposes. The RTP is a 1 MWt pool type reactor with natural convection cooling system and pulsing capability up to 1200 MWt. It went critical on 28 June 1982 and the core configuration has been changed twelve times to date. The core is a mixed type using 20% enriched U-ZrH fuel element containing 8.5, 12 and 20wt% uranium. This paper will discuss the modeling of thermal-hydraulics behaviour in reactor core of RTP using computer code namely PARET. The results of the calculation that were carried out at RTP are modelled and temperature profiles of the thermal hydraulics data at different locations and power levels are developed. s a comparison to the thermal hydraulics calculation using PARET, an experiment were carried out at several different locations and power levels in the reactor core for temperature profile in the core to compare the result obtained from PARET. Finally, an overall analysis of the result of PARET calculation and experimental measurement were exhibited in this paper. (author)

  13. Data from thermal-hydraulic experiments at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Thermal-hydraulic separate effects experimentation in support of nuclear reactor licensing was initiated in the mid-1970s at Oak Ridge National Laboratory (ORNL) and ended in the mid-1980s. A variety of facilities were designed, built, and operated in support of the US Nuclear Regulatory Commission (NRC). These facilities were used to examine a wide variety of thermal-hydraulic phenomena applicable to commercial power reactors. More recently, in the early 1990s, experiments supporting research reactor design were also performed. A summary of the major nuclear reactor experimental efforts is shown in Table I. Each of these experiments generated a significant amount of raw data and also a significant amount of reduced data that have been stored in a variety of ways which are discussed here. Technology has quickly overtaken database archival methods of the past and offers a significant opportunity to improve both the quantity and quality of the thermal-hydraulic information available, while allowing access to a much broader community. Incorporation of future and existing data in a uniform database would greatly benefit principal investigators who access this information

  14. Progress toward Monte Carlo–thermal hydraulic coupling using low-order nonlinear diffusion acceleration methods

    International Nuclear Information System (INIS)

    Highlights: • Method for coupling Monte Carlo to thermal hydraulics using nonlinear acceleration. • Multipole representation of cross sections to perform on-the-fly Doppler broadening. • Support vector machines used for cross section dependence on density and temperature. • Thermal hydraulic feedback is performed during inactive fission source generations. - Abstract: A new approach for coupled Monte Carlo (MC) and thermal hydraulics (TH) simulations is proposed using low-order nonlinear diffusion acceleration methods. This approach uses new features such as coarse mesh finite difference diffusion (CMFD), multipole representation for fuel temperature feedback on microscopic cross sections, and support vector machine learning algorithms (SVM) for iterations between CMFD and TH equations. The multipole representation method showed small differences of about 0.3% root mean square (RMS) error in converged assembly source distribution compared to a conventional MC simulation with ACE data at the same temperature. This is within two standard deviations of the real uncertainty. Eigenvalue differences were on the order of 10 pcm. Support vector machine regression was performed on-the-fly during MC simulations. Regression results of macroscopic cross sections parametrized by coolant density and fuel temperature were successful and eliminated the need of partial derivative tables generated from lattice codes. All of these new tools were integrated together to perform MC–CMFD–TH–SVM iterations. Results showed that inner iterations between CMFD–TH–SVM are needed to obtain a stable solution

  15. Thermal Hydraulic Analysis of 3 MW TRIGA Research Reactor of Bangladesh Considering Different Cycles of Burnup

    Directory of Open Access Journals (Sweden)

    M.H. Altaf

    2014-12-01

    Full Text Available Burnup dependent steady state thermal hydraulic analysis of TRIGA Mark-II research reactor has been carried out utilizing coupled point kinetics, neutronics and thermal hydraulics code EUREKA-2/RR. From the previous calculations of neutronics parameters including percentage burnup of individual fuel elements performed so far for 700 MWD burnt core of TRIGA reactor showed that the fuel rod predicted as hottest at the beginning of cycle (fresh core was found to remain as the hottest until 200 MWD of burn, but, with the progress of core burn, the hottest rod was found to be shifted and another rod in the core became the hottest. The present study intends to evaluate the thermal hydraulic parameters of these hottest fuel rods at different cycles of burnup, from beginning to 700 MWD core burnt considering reactor operates under steady state condition. Peak fuel centerline temperature, maximum cladding and coolant temperatures of the hottest channels were calculated. It revealed that maximum temperature reported for fuel clad and fuel centerline found to lie below their melting points which indicate that there is no chance of burnout on the fuel cladding surface and no blister in the fuel meat throughout the considered cycles of core burnt.

  16. Thermal Hydraulic Characteristic of TRIGA 2000 Reactor for The 110 Percent Normal Power

    International Nuclear Information System (INIS)

    In order to accomplish the Safety Analysis Report (SAR) of TRIGA 2000 reactor according to International Atomic Energy Agency (IAEA) recommendation, the thermal hydraulic aspect’s analysis of TRIGA 2000 reactor for the 110 percent normal power had been carried out by using STAT computer code. STAT code was made by General Atomic and used specifically for analysing the characteristic of TRIGA 2000 reactor. The purpose of the thermal hydraulic analysis is to considerably study the safety problems to achieve thermal hydraulic parameters in the TRIGA 2000 reactor’s core in order to convince that the reactor was not operating unless in the safety condition. Result of this analysis indicated that the film boiling does not occur in the reactor core and DNBR for 2200 kW power with the inlet temperature range between 34 °C – 40 °C is about 2.5 – 2.8 for Mc Adams correlation and about 1.6 – 1.8 for Bernath correlation. (author)

  17. An assessment of the CORCON-MOD3 code. Part 1: Thermal-hydraulic calculations

    International Nuclear Information System (INIS)

    This report deals with the subject of CORCON-Mod3 code validation (thermal-hydraulic modeling capability only) based on MCCI (molten core concrete interaction) experiments conducted under different programs in the past decade. Thermal-hydraulic calculations (i.e., concrete ablation, melt temperature, melt energy, concrete temperature, and condensible and non-condensible gas generation) were performed with the code, and compared with the data from 15 experiments, conducted at different scales using both simulant (metallic and oxidic) and prototypic melt materials, using different concrete types, and with and without an overlying water pool. Sensitivity studies were performed in a few cases involving, for example, heat transfer from melt to concrete, condensed phase chemistry, etc. Further, special analysis was performed using the ACE L8 experimental data to illustrate the differences between the experimental and the reactor conditions, and to demonstrate that with proper corrections made to the code, the calculated results were in better agreement with the experimental data. Generally, in the case of dry cavity and metallic melts, CORCON-Mod3 thermal-hydraulic calculations were in good agreement with the test data. For oxidic melts in a dry cavity, uncertainties in heat transfer models played an important role for two melt configurations--a stratified geometry with segregated metal and oxide layers, and a heterogeneous mixture. Some discrepancies in the gas release data were noted in a few cases

  18. An assessment of the CORCON-MOD3 code. Part 1: Thermal-hydraulic calculations

    Energy Technology Data Exchange (ETDEWEB)

    Strizhov, V.; Kanukova, V.; Vinogradova, T.; Askenov, E. [Russian Academy of Sciences, Moscow (Russian Federation). Inst. of Nuclear Safety; Nikulshin, V. [Russian Research Center, Moscow (Russian Federation). Kurchatov Inst.

    1996-09-01

    This report deals with the subject of CORCON-Mod3 code validation (thermal-hydraulic modeling capability only) based on MCCI (molten core concrete interaction) experiments conducted under different programs in the past decade. Thermal-hydraulic calculations (i.e., concrete ablation, melt temperature, melt energy, concrete temperature, and condensible and non-condensible gas generation) were performed with the code, and compared with the data from 15 experiments, conducted at different scales using both simulant (metallic and oxidic) and prototypic melt materials, using different concrete types, and with and without an overlying water pool. Sensitivity studies were performed in a few cases involving, for example, heat transfer from melt to concrete, condensed phase chemistry, etc. Further, special analysis was performed using the ACE L8 experimental data to illustrate the differences between the experimental and the reactor conditions, and to demonstrate that with proper corrections made to the code, the calculated results were in better agreement with the experimental data. Generally, in the case of dry cavity and metallic melts, CORCON-Mod3 thermal-hydraulic calculations were in good agreement with the test data. For oxidic melts in a dry cavity, uncertainties in heat transfer models played an important role for two melt configurations--a stratified geometry with segregated metal and oxide layers, and a heterogeneous mixture. Some discrepancies in the gas release data were noted in a few cases.

  19. PATHS: a steady state two-phase thermal hydraulic solver for PARCS depletion

    International Nuclear Information System (INIS)

    The PATHS code was developed to solve the steady state two-phase thermal-hydraulic equations for a Boiling Water Reactor (BWR) and to provide thermal-hydraulic feedback for BWR depletion calculations with the neutronics code PARCS. The PARCS code is coupled to RELAP5 and TRACE which are normally used to solve for the thermal hydraulic state for BWR applications. However, systems codes were developed primarily for transient analysis and it can be computationally expensive to perform null transients to achieve the steady-state for the many channel problems required for practical BWR depletion analysis. For steady state analysis of the reactor, it is much more efficient to use a lower order two phase solution methodology. The low order methodology improves the runtime without major compromises in the fluid density and temperature distributions that are important for depletion analysis. In the PATHS code, the drift flux model is used with the EPRI void model. PATHS results were compared to TRACE for fixed power computations at various powers and flow rates. Coupled PATHS/PARCS calculations were then validated using depletion data from cycles 1 and 2 of the Peach Bottom II BWR. (author)

  20. Advanced materials for thermal protection system

    Science.gov (United States)

    Heng, Sangvavann; Sherman, Andrew J.

    1996-03-01

    Reticulated open-cell ceramic foams (both vitreous carbon and silicon carbide) and ceramic composites (SiC-based, both monolithic and fiber-reinforced) were evaluated as candidate materials for use in a heat shield sandwich panel design as an advanced thermal protection system (TPS) for unmanned single-use hypersonic reentry vehicles. These materials were fabricated by chemical vapor deposition/infiltration (CVD/CVI) and evaluated extensively for their mechanical, thermal, and erosion/ablation performance. In the TPS, the ceramic foams were used as a structural core providing thermal insulation and mechanical load distribution, while the ceramic composites were used as facesheets providing resistance to aerodynamic, shear, and erosive forces. Tensile, compressive, and shear strength, elastic and shear modulus, fracture toughness, Poisson's ratio, and thermal conductivity were measured for the ceramic foams, while arcjet testing was conducted on the ceramic composites at heat flux levels up to 5.90 MW/m2 (520 Btu/ft2ṡsec). Two prototype test articles were fabricated and subjected to arcjet testing at heat flux levels of 1.70-3.40 MW/m2 (150-300 Btu/ft2ṡsec) under simulated reentry trajectories.

  1. Predicting Formation Damage in Aquifer Thermal Energy Storage Systems Utilizing a Coupled Hydraulic-Thermal-Chemical Reservoir Model

    Science.gov (United States)

    Müller, Daniel; Regenspurg, Simona; Milsch, Harald; Blöcher, Guido; Kranz, Stefan; Saadat, Ali

    2014-05-01

    In aquifer thermal energy storage (ATES) systems, large amounts of energy can be stored by injecting hot water into deep or intermediate aquifers. In a seasonal production-injection cycle, water is circulated through a system comprising the porous aquifer, a production well, a heat exchanger and an injection well. This process involves large temperature and pressure differences, which shift chemical equilibria and introduce or amplify mechanical processes. Rock-fluid interaction such as dissolution and precipitation or migration and deposition of fine particles will affect the hydraulic properties of the porous medium and may lead to irreversible formation damage. In consequence, these processes determine the long-term performance of the ATES system and need to be predicted to ensure the reliability of the system. However, high temperature and pressure gradients and dynamic feedback cycles pose challenges on predicting the influence of the relevant processes. Within this study, a reservoir model comprising a coupled hydraulic-thermal-chemical simulation was developed based on an ATES demonstration project located in the city of Berlin, Germany. The structural model was created with Petrel, based on data available from seismic cross-sections and wellbores. The reservoir simulation was realized by combining the capabilities of multiple simulation tools. For the reactive transport model, COMSOL Multiphysics (hydraulic-thermal) and PHREEQC (chemical) were combined using the novel interface COMSOL_PHREEQC, developed by Wissmeier & Barry (2011). It provides a MATLAB-based coupling interface between both programs. Compared to using COMSOL's built-in reactive transport simulator, PHREEQC additionally calculates adsorption and reaction kinetics and allows the selection of different activity coefficient models in the database. The presented simulation tool will be able to predict the most important aspects of hydraulic, thermal and chemical transport processes relevant to

  2. Influence of hydraulics and control of thermal storage in solar assisted heat pump combisystems

    OpenAIRE

    Poppi, Stefano; Bales, Chris

    2014-01-01

    This paper studies the influence of hydraulics and control of thermal storage in systems combined with solar thermal and heat pump for the production of warm water and space heating in dwellings. A reference air source heat pump system with flat plate collectors connected to a combistore was defined and modeled together with the IEA SHC Task 44 / HPP Annex 38 (T44A38) “Solar and Heat Pump Systems” boundary conditions of Strasbourg climate and SFH45 building. Three and four pipe connections as...

  3. Advanced model structures applied to system identification of a servo- hydraulic test rig

    Directory of Open Access Journals (Sweden)

    P. Czop

    2010-07-01

    Full Text Available Purpose: This paper deals with a method for the parametric system identification of a nonlinear system to obtain its parametric representation using a linear transfer function. Such representation is applicable in off-line profile correction methods minimizing the error between a reference input signal and a signal performed by the test rig. In turn, a test signal can be perfectly tracked by a servo-hydraulic test rig. This is the requirement in massive production where short test sequences are repeated to validate the products.Design/methodology/approach: A numerical and experimental case studies are presented in the paper. The numerical study presents a system identification process of a nonlinear system consisting of a linear transfer function and a nonlinear output component, being a static function. The experimental study presents a system identification process of a nonlinear system which is a servo-hydraulic test rig. The simulation data has been used to illustrate the feasibility study of the proposed approach, while the experimental data have been used to validate advanced model structures under operational conditions.Findings: The advanced model structures confirmed their better performance by means of the model fit in the time domain.Research limitations/implications: The method applies to analysis of such mechanical and hydraulic systems for which measurements are corrupted by residual harmonic disturbances resulting from system nonlinearities.Practical implications: The advanced model structures are intended to be used as inverse models in off-line signal profile correction.Originality/value: The results state the foundation for the off-line parametric error cancellation method which aims in improving tracking of load signals on servo-hydraulic test rigs.

  4. Uncertainty Assessment of the Core Thermal-Hydraulic Analysis Using the Monte Carlo Method

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Sun Rock; Yoo, Jae Woon; Hwang, Dae Hyun; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    In the core thermal-hydraulic design of a sodium cooled fast reactor, the uncertainty factor analysis is a critical issue in order to assure safe and reliable operation. The deviations from the nominal values need to be quantitatively considered by statistical thermal design methods. The hot channel factors (HCF) were employed to evaluate the uncertainty in the early design such as the CRBRP. The improved thermal design procedure (ISTP) calculates the overall uncertainty based on the Root Sum Square technique and sensitivity analyses of each design parameters. Another way to consider the uncertainties is to use the Monte Carlo method (MCM). In this method, all the input uncertainties are randomly sampled according to their probability density functions and the resulting distribution for the output quantity is analyzed. It is able to directly estimate the uncertainty effects and propagation characteristics for the present thermalhydraulic model. However, it requires a huge computation time to get a reliable result because the accuracy is dependent on the sampling size. In this paper, the analysis of uncertainty factors using the Monte Carlo method is described. As a benchmark model, the ORNL 19 pin test is employed to validate the current uncertainty analysis method. The thermal-hydraulic calculation is conducted using the MATRA-LMR program which was developed at KAERI based on the subchannel approach. The results are compared with those of the hot channel factors and the improved thermal design procedure

  5. RELAP5-3D thermal hydraulic analysis of the target cooling system in the SPES experimental facility

    International Nuclear Information System (INIS)

    The SPES (Selective Production of Exotic Species) experimental facility, under construction at the Italian National Institute of Nuclear Physics (INFN) Laboratories of Legnaro, Italy, is a second generation Isotope Separation On Line (ISOL) plant for advanced nuclear physic studies. The UCx target-ion source system works at temperature of about 2273 K, producing a high level of radiation (105 Sv/h), for this reason a careful risk analysis for the target chamber is among the major safety issues. In this paper, the obtained results of thermofluid-dynamics simulations of accidental transients in the SPES target cooling system are reported. The analysis, performed by using the RELAP5-3D 2.4.2 qualified thermal-hydraulic system code, proves good safety performance of this system during different accidental conditions

  6. Simulation of the passive condensation cooling tank of the PASCAL test facility using the component thermal-hydraulic analysis code CUPID

    International Nuclear Information System (INIS)

    For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal hydraulics code, named CUPID, has been being developed. In the present study, the CUPID code was applied for the simulation of the PASCAL (PAFS Condensing Heat Removal Assessment Loop) test facility constructed with an aim of validating the cooling and operational performance of the PAFS (Passive Auxiliary Feedwater System). The PAFS is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor +), which is intended to completely replace the conventional active auxiliary feedwater system. This paper presents the preliminary simulation results of the PASCAL facility performed with the CUPID code in order to verify its applicability to the thermal-hydraulic phenomena inside the system. A standalone calculation for the passive condensation cooling tank was performed by imposing a heat source boundary condition and the transient thermal-hydraulic behaviors inside the system, such as the water level, temperature and velocity, were qualitatively investigated. The simulation results verified that the natural circulation and boiling phenomena in the water pool can be well reproduced by the CUPID code. (authors)

  7. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  8. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Science.gov (United States)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-01

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  9. Nuclear fuel element design and thermal-hydraulic analysis of Wolsung-1, 600 MWe CANDU-PHWR (Part II)

    International Nuclear Information System (INIS)

    The main objective of the present thermal hydraulic analysis is to determine the thermal hydraulic characteristics of Wolsung-1 600 MWe CANDU-PHW reactor under normal operation. This is to verify and expedite the development of the nuclear fuel design and fabrication as well as the management. The computer program package developed for the stated objective are DOD81, CANREPP, PLOC81 and COBRA-CANDU. (Author)

  10. Whole Core Pin-by-Pin Coupled Neutronic-Thermal-hydraulic Steady state and Transient Calculations using COBAYA3 code

    OpenAIRE

    Jiménez Escalante, Javier; Herrero Carrascosa, José Javier; Cuervo Gómez, Diana; Aragonés Beltrán, José María

    2010-01-01

    Nowadays, coupled 3D neutron-kinetics and thermal-hydraulic core calculations are performed by applying a radial average channel approach using a meshing of one quarter of assembly in the best case. This approach does not take into account the subchannels effects due to the averaging of the physical fields and the loose of heterogeneity in the thermal-hydraulic modelization. Therefore the models do not have enough resolution to predict those subchannels effects which are important for the fue...

  11. Thermal-hydraulic safety analyses supporting the steam generator replacement and uprating at Krško nuclear power plant

    OpenAIRE

    Mavko, Borut; Prošek, Andrej

    2015-01-01

    The Krško nuclear power plant has undertaken a major modernization project. The objectives of the project are: long-term stabilization of the plant's operation, uprating of the net electrical power output, higher availability and enhanced safety of the plant. The modernization also requires a thorough safety re-avaluation and therefore new thermal hydraulic, mechanical and structural analysis. The thermal-hydraulic part of the safety analysis necessary for the steam generator replacement and ...

  12. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study

  13. Thermal-hydraulic simulation and analysis of Research Reactor Cooling Systems

    International Nuclear Information System (INIS)

    The objective of the present study is to formulate a model to simulate the thermal hydraulic behavior of integrated cooling system in a typical material testing reactor (MTR) under loss of ultimate heat sink, the model involves three interactively coupled sub-models for reactor core, heat exchanger and cooling tower. The developed model predicts the temperature profiles in addition it predicts inlet and outlet temperatures of the hot and cold stream as well as the heat exchangers and cooling tower. The model is validated against PARET code for steady-state operation and also verified by the reactor operational records, and then the model is used to simulate the thermal-hydraulic behavior of the reactor under a loss of ultimate heat sink. The simulation is performed for two operational regimes named regime I of (11 MW) thermal power and three operated cooling tower cells and regime II of (22 MW) thermal power and six operated cooling tower cells. In regime I, the simulation is performed for 1, 2 and 3 cooling tower failed cells while in regime II, it is performed for 1, 2, 3, 4, 5 and 6 cooling tower failed cells. The safety action is conducted by the reactor protection system (RPS) named power reduction safety action, it is triggered to decrease the reactor power by amount of 20% of the present power when the water inlet temperature to the core reaches 43 degree C and a scram (emergency shutdown) is triggered in case of the inlet temperature reaches 44 degree C. The model results are analyzed and discussed. The temperature profiles of fuel, clad and coolant are predicted during transient where its maximum values are far from thermal hydraulic limits.

  14. Thermal-hydraulic performance and structural thermal stress analysis for ITER shield blanket module nearby NB rejoin

    International Nuclear Information System (INIS)

    Hydraulic and thermal analysis of the International Thermonuclear Experimental Reactor (ITER) standard neutral beam (NB) blanket module was carried out in order to check whether the latest design meets ITER requirements. Minor-loss coefficients were estimated with a CFD code, and friction factors of straight channels were obtained using existing formulas. The effects of different radial hole's diameter, length of the back of the radial hole, size of clearance, type of flow driver, branch velocity and flow direction on minor-loss coefficients for radial holes were investigated. Since total mass flow rate and dimensions of the cooling channels were given, when pressure drop due to intersection of the radial hole with back drilled collector was ignored, we can obtain pressure drop, flow rate, velocity and heat transfer coefficient in each radial hole. An improved calculation without neglecting the pressure drop caused by the intersection was also done to compare with the simplified one. Finally, maximum temperature, thermal stress and deformation were evaluated according to FEM thermal analysis. The results of the latest hydraulic and thermal analysis indicate that the current design meets ITER requirements well, except that flow distribution is not so uniform when different types of flow drivers are used, and temperature in the front head surface is a little high. Improved design is necessary in the further. (authors)

  15. Use of a hydraulic brake as a source of thermal energy for the railway rolling stock

    Directory of Open Access Journals (Sweden)

    V.A.Gabrinets

    2012-12-01

    Full Text Available Introduction: In this paper the braking issues of passenger trains which have a great speed and frequent stops are examines. Problem statement: These processes are ехpensive and have big energy losses. The proposed solution to the problem: The kinetic energy of braking prosses propose to turn into thermal energy of heating fluid. For this purpose special hydraulic brake is proposed. The brake is connected with the wheel carriage pairs. The process is based on the energy dissipation in liqid when the disks with spikes rotate in it. Because the real liquid has friction and viscosity, it will be heat up, when the mechanical parts of the hydraulic brake are moved in it. The design, operating principle and characteristics of the hydraulic brake are proposed. Transmission of kinetic energy of carriage motion to brake system executed by mechanical clutches. It connected with the wheel pair and transmitting the energy the wheels rotation to hydraulic brake discs. The cylindrical rods are installed on the discs. Rods location fits the profile of the curved centrifugal pump vanes. As result, the fluid heatind prosess by rotatinge discs with rods take place also at the same time with the liquid pumping through the inner volume of brake system.Conclusions: Affordable passenger carriage braking dynamic is achieved by varying the size and number of rods. The heated liquid may be subsequently used for household needs and for heating the passenger carriage.

  16. Comparative analysis of CTF and trace thermal-hydraulic codes using OECD/NRC PSBT benchmark void distribution database

    International Nuclear Information System (INIS)

    The international OECD/NRC PWR Subchannel and Bundle Tests (PSBT) benchmark has been established to provide a test bed for assessing the capabilities of various thermal-hydraulic subchannel, system, and computational fluid dynamics (CFD) codes and to encourage advancement in the analysis of fluid flow in rod bundles. The aim is to improve the reliability of the nuclear reactor safety margin evaluations. The benchmark is based on one of the most valuable databases identified for the thermal-hydraulics modeling, which was developed by the Nuclear Power Engineering Corporation (NUPEC) in Japan. The database includes subchannel void fraction and departure from nucleate boiling (DNB) measurements in a representative Pressurized Water Reactor (PWR) fuel assembly. Part of this database is made available for the international PSBT benchmark activity. The PSBT benchmark team is organized based on the collaboration between the Pennsylvania State University (PSU) and the Japan Nuclear Energy Safety organization (JNES) including the participation and support of the U.S. Nuclear Regulatory Commission (NRC) and the Nuclear Energy Agency (NEA), OECD. On behalf of the PSBT benchmark team, PSU in collaboration with US NRC is performing supporting calculations of the benchmark exercises using its in-house advanced thermalhydraulic subchannel code CTF and the US NRC system code TRACE. CTF is a version of the well-known and widely used code COBRA-TF whose models have been continuously improved and validated over the last years at the Reactor Dynamics and Fuel Management Group (RDFMG) at PSU. TRACE is a reactor systems code developed by the U.S. Nuclear Regulatory Commission to analyze transient and steady-state thermal-hydraulic behavior in Light Water Reactors (LWRs) and it has been designed to perform best-estimate analyses of loss-of-coolant accidents (LOCAs), operational transients, and other accident scenarios in PWRs and boiling light-water reactors (BWRs). The paper presents

  17. IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bostelmann, F. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained). SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on

  18. IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

    International Nuclear Information System (INIS)

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained). SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on

  19. Thermal-hydraulic behavior in downcomer during refill phase of LBLOCA with downcomer injection

    International Nuclear Information System (INIS)

    The primary objective of this study is assessment of predictability of two codes, TRAC-PF1/ MOD2 and RELAP5/MOD3, which are known as currently available best-estimate codes in predicting the major thermal-hydraulic phenomena during refill phase of LBLOCA with downcomer injection. It is known that TRAC code has three-dimensional hydraulic capability and that the RELAP5/MOD3 has well-qualified two-phase constitutive models. The secondary objective is to recommend the considerations including modeling scheme in calculating the real plant ECCS performance based on the present assessment. For those purpose, one of the Upper Plenum Test Facility (UPTF) tests, which showed a typical three-dimensional behavior in downcomer injection of ECCS, was simulated by two codes. System pressure response, break flow, and lower plenum level were compared with the applicable experimental data and analyzed in terms of the major thermal hydraulic phenomena, ECC bypass, steam sweep out in lower plenum and downcomer etc. To provide modeling scheme considerations, effects of vessel modeling scheme, downcomer wall friction and ECC injection point modeling were investigated as sensitivity study. (author)

  20. Effect of the discretization and neutronic thermal hydraulic coupling on LWR transients

    International Nuclear Information System (INIS)

    Coupled multiphysics problems solve different physical phenomena with time scales of varying orders of magnitude. These phenomena are coupled in a nonlinear way making it difficult to find an accurate and efficient solution. Many of the present generation of codes for LWR are based on 3-D neutronic nodal methods coupled with first order thermal hydraulic methods. Moreover, the spatial and temporal meshes used to solve each field are different reflecting the scales of each phenomenon. This paper discusses the effect of the spatial and temporal discretization as well as the effect of different coupling schemes, with different level of implicitness, between the neutronic and core thermal hydraulics in SIMULATE-3K (S3K). S3K is a best estimate code used by many utilities, regulatory authorities and research institutes for the analysis of LWR transients that require the coupling of neutronic, fuel pin, and core hydraulic models. Examples of S3K applications are BWR stability analysis, fast anticipated operational occurrences, with or without scram, and reactivity initiated transients. Three different applications will be discussed in this paper to illustrate the effect of the discretization and coupling methods in multiphysics problems, namely: the NEA PWR rod ejection, the Ringhals-1 BWR stability, and the Peach Bottom turbine trip benchmarks. (author)

  1. Thermal and hydraulic test plan of TRU fuel element for transmutation process

    International Nuclear Information System (INIS)

    JAERI is developing processes to partition long-lived transuranic elements (TRU) from high-level radioactive waste and transmutation processes to transform TRU into shorter-lived or stable nuclides under the OMEGA program. To promote developments of transmutation processes, thermal and hydraulic tests were planed to optimize a fuel element of an actinide burner fast reactor (ABR) cooled by helium gas. Along the test plan, a simulated fuel element in which simulated fuel particles were filled up in the porous annular space of 11.7mm in gap width and of 600mm in length was manufactured experimentally, and also a test apparatus which could circulate helium gas or nitrogen gas at a maximum flow rate of 400 m3/h under 1 MPa was designed and fabricated. Hydraulic performance of the test apparatus was confirmed through preliminary operations. This paper presents mainly a thermal and hydraulic test plan of the fuel element for developing ABR core design, outlines of the simulated fuel element and the test apparatus, and preliminary operation results. (author)

  2. Fundamental study on thermo-hydraulic phenomena concerning passive safety of advanced marine reactor

    International Nuclear Information System (INIS)

    The objective of this study is to investigate the thermo-hydraulic behavior of a fluid region confined in a rectangular parallelepiped cavity equipped with a heater and a cooler. The motivation of this study is to clarify a thermal buffer effect for an innovative marine nuclear reactor to realize passive safety. In the present study, experiments were carried out with conditions of laminar convection. Temperature and flow behavior was visualized by the liquid-crystal suspension method, by which the temperature distribution in liquid can be observed as a colored map. Thermal plumes from the heater and the cooler, global natural circulation in the cavity and thermal stratification were observed as elements of the complicated phenomena. Using a code which solves the Navier-Stokes and energy equations, numerical simulations under steady and unsteady condition were carried out to predict the experimental results for two-dimensional, laminar situations, and a good agreement was obtained. (author)

  3. Sensitivity analysis of hydraulic and thermal parameters inducing anomalous heat flow in the Lower Yarmouk Gorge

    Science.gov (United States)

    Goretzki, Nora; Inbar, Nimrod; Kühn, Michael; Möller, Peter; Rosenthal, Eliyahu; Schneider, Michael; Siebert, Christian; Magri, Fabien

    2016-04-01

    The Lower Yarmouk Gorge, at the border between Israel and Jordan, is characterized by an anomalous temperature gradient of 46 °C/km. Numerical simulations of thermally-driven flow show that ascending thermal waters are the result of mixed convection, i.e. the interaction between the regional flow from the surrounding heights and buoyant flow within permeable faults [1]. Those models were calibrated against available temperature logs by running several forward problems (FP), with a classic "trial and error" method. In the present study, inverse problems (IP) are applied to find alternative parameter distributions that also lead to the observed thermal anomalies. The investigated physical parameters are hydraulic conductivity and thermal conductivity. To solve the IP, the PEST® code [2] is applied via the graphical interface FEPEST® in FEFLOW® [3]. The results show that both hydraulic and thermal conductivity are consistent with the values determined with the trial and error calibrations, which precede this study. However, the IP indicates that the hydraulic conductivity of the Senonian Paleocene aquitard can be 8.54*10-3 m/d, which is three times lower than the originally estimated value in [1]. Moreover, the IP suggests that the hydraulic conductivity in the faults can increase locally up to 0.17 m/d. These highly permeable areas can be interpreted as local damage zones at the faults/units intersections. They can act as lateral pathways in the deep aquifers that allow deep outflow of thermal water. This presentation provides an example about the application of FP and IP to infer a wide range of parameter values that reproduce observed environmental issues. [1] Magri F, Inbar N, Siebert C, Rosenthal E, Guttman J, Möller P (2015) Transient simulations of large-scale hydrogeological processes causing temperature and salinity anomalies in the Tiberias Basin. Journal of Hydrology, 520, 342-355 [2] Doherty J (2010) PEST: Model-Independent Parameter Estimation. user

  4. Thermal hydraulics in the hot pool of Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Sodium cooled Fast Breeder Test Reactor (FBTR) of 40 MWt/13 MWe capacity is in operation at Kalpakkam, near Chennai. Presently it is operating with a core of 10.5 MWt. Knowledge of temperatures and flow pattern in the hot pool of FBTR is essential to assess the thermal stresses in the hot pool. While theoretical analysis of the hot pool has been conducted by a three-dimensional code to access the temperature profile, it involves tuning due to complex geometry, thermal stresses and vibration. With this in view, an experimental model was fabricated in 1/4 scale using acrylic material and tests were conducted in water. Initially hydraulic studies were conducted with ambient water maintaining Froude number similarity. After that thermal studies were conducted using hot and cold water maintaining Richardson similitude. In both cases Euler similarity was also maintained. Studies were conducted simulating both low and full power operating conditions. This paper discusses the model simulation, similarity criteria, the various thermal hydraulic studies that were carried out, the results obtained and the comparison with the prototype measurements.

  5. Monte Carlo Neutronics and Thermal Hydraulics Analysis of Reactor Cores with Multilevel Grids

    Science.gov (United States)

    Bernnat, W.; Mattes, M.; Guilliard, N.; Lapins, J.; Zwermann, W.; Pasichnyk, I.; Velkov, K.

    2014-06-01

    Power reactors are composed of assemblies with fuel pin lattices or other repeated structures with several grid levels, which can be modeled in detail by Monte Carlo neutronics codes such as MCNP6 using corresponding lattice options, even for large cores. Except for fresh cores at beginning of life, there is a varying material distribution due to burnup in the different fuel pins. Additionally, for power states the fuel and moderator temperatures and moderator densities vary according to the power distribution and cooling conditions. Therefore, a coupling of the neutronics code with a thermal hydraulics code is necessary. Depending on the level of detail of the analysis, a very large number of cells with different materials and temperatures must be regarded. The assignment of different material properties to all elements of a multilevel grid is very elaborate and may exceed program limits if the standard input procedure is used. Therefore, an internal assignment is used which overrides uniform input parameters. The temperature dependency of continuous energy cross sections, probability tables for the unresolved resonance region and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. The method is applied with MCNP6 and proven for several full core reactor models. For the coupling of MCNP6 with thermal hydraulics appropriate interfaces were developed for the GRS system code ATHLET for liquid coolant and the IKE thermal hydraulics code ATTICA-3D for gaseous coolant. Examples will be shown for different applications for PWRs with square and hexagonal lattices, fast reactors (SFR) with hexagonal lattices and HTRs with pebble bed and prismatic lattices.

  6. Contribution to the study of thermal-hydraulic problems in nuclear reactors

    International Nuclear Information System (INIS)

    In nuclear reactors, whatever the type considered, Pressurized Water Water Reactors (PWRs), Fast Breeder reactors (FBRs)..., thermal-hydraulics, the science of fluid mechanics and thermal behaviour, plays an essential role, both in nominal operating and accidental conditions. Fluid can either be the primary fluid (liquid or gas) or a very specific fluid called corium, which, in case of severe accident, could result from core and environning structure melting. The work reported here represents a 20-year contribution to thermal-hydraulic issues which could occur in FBRs and PWRs. Working on these two types of reactors, both in nominal and severe accident situations, has allowed me to compare the problems and to realize the importance of communication between research teams. The evolution in the complexity of studied problems, unavoidable in order to reduce costs and significantly improve safety, has led me from numerical modelling of single-phase flow turbulence to high temperature real melt experiments. The difficulties encountered in understanding the observed phenomena and in increasing experimental databases for computer code qualification have often entailed my participation in specific measurement device developments or adaptations, in particular non-intrusive devices generally based on optical techniques. Being concerned about the end-use of this research work, I actively participated in 'in-situ' thermalhydraulic experiments in the FBRs: Phenix and Super-Phenix, of which I appreciated their undeniable scientific contribution. In my opinion, the thermal-hydraulic questions related to severe accidents are the most complex as they are at the cross-roads of several scientific specialities. Consequently, they require a multi-disciplinary approach and a continuous see-saw motion between experimentalists and modelling teams. After a brief description of the various problems encountered, the main ones are reported. Finally, the importance for research teams to

  7. Thermal, hydraulic, and mechanical initial conditions around KAERI Underground Research Tunnel

    International Nuclear Information System (INIS)

    In KAERI underground research tunnel(KURT) various in situ experiments for the investigation of thermal, mechanical, hydraulic, and chemical behaviours related to the validation of high-level radioactive waste disposal system are carrying out. In this study, the geological characteristics, thermal, hydraulic, and mechanical(THM) properties of the rock mass, and groundwater level analyzed and derived relationship between the THM properties and depth. From this study, it was found that the THM properties varies with depth Z and many properties could be expressed well with an equation type, a+b/Zc. The calculated rock thermal properties were 3∼7% higher than the measurement and the difference was relatively higher in dry rock. With empirical equations and measured air and tunnel wall temperatures, it was also possible to estimate that the seasonal temperature variations at 5m and 10m distance from tunnel wall were 3 .deg. C and 0,75 .deg. C, respectively. The thermal-hydraulic-mechanical initial conditions around KURT derived from this study will be utilized for the selection of location and the design for various in situ experiments at KURT. Those will be also used as fundamental data for the analysis of the results from the in situ experiments. The understanding of the THM initial conditions will be useful for the investigation of low and intermediate level repository as well the site selection and system design for a temporary storage facility and a high-level radioactive waste repository. This will also be applied to the design of underground caverns for various purposes and the analysis of in situ measurements at underground excavations

  8. Thermal-Hydraulic Assessment Of HLM-Cooled Pin Bundle In CIRCE Pool Facility

    International Nuclear Information System (INIS)

    Since the Lead-cooled Fast Reactor (LFR) has been conceptualized in the frame of GEN IV International Forum (GIF), ENEA is strongly involved on the HLM technology development. Currently ENEA has implemented large competencies and capabilities in the field of HLM thermal-hydraulic, coolant technology, material for high temperature applications, corrosion and material protection, heat transfer and removal, component development and testing, remote maintenance, procedure definition and coolant handling. In this frame the CIRCE pool facility has been refurbished to host a suitable test section able to thermal-hydraulically simulate the primary system of a HLM cooled pool reactor. In particular a fuel pin bundle simulator (FPS) has been installed in the CIRCE pool. It has been conceived with a thermal power of about 1 MW and a linear power up to 25 kW/m, relevant values for a LMFR. It consist of 37 fuel pins (electrically simulated) placed on a hexagonal lattice with a pitch to diameter ratio of 1.8. The pins have a diameter of 8.2 mm and an active length of 1 m. Along the FPS, three spacer grid properly designed by ENEA have been installed. The FPS has been deeply instrumented by several thermocouples. In particular three sections of the FPS have been instrumented to monitor the heat transfer coefficient along the bundle as well as the cladding temperature in different rank of sub-channels. A first set of experiments were run to investigate the thermal-hydraulic behavior of the fuel pin bundle both under forced (by gas lift) and buoyancy driven circulation into the pool. The paper reports the experimental data carried out as well as a preliminary analysis and discussion, also in comparison with CFD calculations performed by CFX code. (author)

  9. A comparison of the CHF between tubes and annuli under PWR thermal-hydraulic conditions

    Energy Technology Data Exchange (ETDEWEB)

    Herer, C. [RRAMATOME EP/TC, Paris (France); Souyri, A. [EdF DER/RNE/TTA, Chatou (France); Garnier, J. [CEA DRN/DTP/STR/LETC, Grenoble (France)

    1995-09-01

    Critical Heat Flux (CHF) tests were carried out in three tubes with inside diameters of 8, 13, and 19.2 mm and in two annuli with an inner tube of 9.5 mm and an outer tube of 13 or 19.2 mm. All axial heat flux distributions in the test sections were uniform. The coolant fluid was Refrigerant 12 (Freon-12) under PWR thermal-hydraulic conditions (equivalent water conditions - Pressure: 7 to 20 MPa, Mass Velocity: 1000 to 6000 kg/m2/s, Local Quality: -75% to +45%). The effect of tube diameter is correlated for qualities under 15%. The change from the tube to the annulus configuration is correctly taken into account by the equivalent hydraulic diameter. Useful information is also provided concerning the effect of a cold wall in an annulus.

  10. Development and test of interphase friction model for reactor thermal-hydraulic safety analysis

    International Nuclear Information System (INIS)

    Background: In order to change the status that nuclear power software technology lags behind the foreign countries, the thermal-hydraulic safety analysis program COSINE is being developed. Purpose: To complete the solution of the hydraulics equation, a fully considered model of detailed vapor and liquid interphase friction was proposed. Methods: With reference to some mature interphase friction models, we developed the FORTRAN program for the model and designed appropriate numerical routines to test the model. And the RELAP5/Mod3.2 was applied to results comparison of the rationality and accuracy of the model. Results and Conclusion: The results show that the model is valid and the program can run normally. (authors)

  11. Thermal-hydraulic modeling and analysis of spool valve with sloping U-shape notch by bond graph

    Institute of Scientific and Technical Information of China (English)

    娄磊; 吴万荣; 王兆强; 梁向京

    2015-01-01

    To increase the efficiency and reliability of the thermodynamics analysis of the spool valve, the precise function expression of the flow area for the sloping U-shape notch orifice versus the spool stroke and thermal-hydraulic bond graph based on the conservation of mass and energy were introduced. Subsequently, the connection rule for the bond graph elements and the method to construct the complete thermal-hydraulic system model were proposed. On the basis of heat transfer analysis of a typical hydraulic circuit containing the spool valve, the lumped parameter for mathematical model of the system was given. At last, the reliability of the mathematical model of the flow area and the thermal-hydraulic system for the sloping U-shape notch orifice on the spool were demonstrated by the test. The good agreement between the simulation results and experimental data demonstrates the validity of the modeling method.

  12. Evaluation on numerical simulation accuracy of the commercial CFD program for FBR thermal-hydraulic conditions and applications. Single phase multi-dimensional thermal-hydraulic evaluation problems

    International Nuclear Information System (INIS)

    Commercial computational fluid dynamic program is taken up to be employed for nuclear thermal-hydraulic applications due to the advantages in high-speed solution and easy-to-use operation. The principal objective of this report is evaluating the numerical simulation accuracy of the Fluent, on single-phase multi-dimensional thermal hydraulic problems. The evaluation problems are: 1) Laminar flow over a backward-facing step, 2) Turbulent flow over a backward-facing step, 3) Temperature of a inner rectangular rotating flow, 4) Thermal-driven natural convection flow in a square cavity, and 5) Turbulent flow in a cubic cavity, those were selected in supposing nuclear reactor thermal-hydraulic conditions by the technical committee of the Japan atomic energy society. The features on numerical method and accuracy of the Fluent being identified are: 1) Spatial differential schemes for convection term: 1st upwind, power-law, 2nd upwind, and Quick, upgrade the numerical accuracy in this order. Each scheme has the same accuracy as of the existing referenced numerical results. Quick scheme employs numerical stability oriented filtering so that no over- or under-shoots are observed. Yet, 2nd central differential scheme -used in large eddy simulation (LES)- leads numerical instability (i.e. temporal oscillation in pressure, and spatial wavering in velocity) typically when we deal with in low-resolution domains. 2) Turbulent models: (Standard, RNG, Realizable) k-ε, (Standard, SST) k-ω, and, (Standard, Quadratic) RST, necessitate to involve non-equilibrium wall function to take numerical accuracy and stability. The Fluent evaluations on re-attaching points and velocity distributions show nearly the same as -and on several counts more accurate than- those of the existing reference results. The LES turbulent model can be used only for 3-D simulations. 3) The evaluations of thermal-driven natural convection flow, which is one of the heat transfer and fluidics coupling problem, show

  13. High temperature gas-cooled pebble bed reactor steady state thermal-hydraulics analyses based on CFD method

    International Nuclear Information System (INIS)

    Background: Based on general purpose CFD code Fluent, the PBMR-400 full load nominal condition thermal-hydraulics performance was studied by applying local thermal non-equilibrium porous media model. Purpose: In thermal hydraulics study of the gas cooled pebble bed reactor, the core of the reactor can be treated as macroscopic porous media with strong inner heat source, and the original Fluent code can not handle it properly. Methods: By introducing a UDS in the calculation domain of the reactor core and subjoining a new resistance term, we develop a non-equilibrium porous media model which can give an accurate description of the core of the pebble bed. The mesh of CFD code is finer than that of the traditional pebble bed reactor thermal hydraulics analysis code such as THERMIX and TINTE, thus more information about coolant velocity fields, temperature field and solid phase temperature field can be acquired. Results: The nominal condition calculation results of the CFD code are compared to those of the well-established thermal-hydraulic code THERMIX and TINTE, and show a good consistency. Conclusion: The extended local thermal non-equilibrium model can be used to analyse thermal-hydraulics of high temperature pebble bed type reactor. (authors)

  14. PEBBLE: a two-dimensional steady-state pebble bed reactor thermal hydraulics code

    International Nuclear Information System (INIS)

    This report documents the local implementation of the PEBBLE code to treat the two-dimensional steady-state pebble bed reactor thermal hydraulics problem. This code is implemented as a module of a computation system used for reactor core history calculations. Given power density data, the geometric description in (RZ), and basic heat removal conditions and thermal properties, the coolant properties, flow conditions, and temperature distributions in the pebble fuel elements are predicted. The calculation is oriented to the continuous fueling, steady state condition with consideration of the effect of the high energy neutron flux exposure and temperature history on the thermal conductivity. The coolant flow conditions are calculated for the same geometry as used in the neutronics calculation, power density and fluence data being used directly, and temperature results are made available for subsequent use

  15. PEBBLE: a two-dimensional steady-state pebble bed reactor thermal hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1981-09-01

    This report documents the local implementation of the PEBBLE code to treat the two-dimensional steady-state pebble bed reactor thermal hydraulics problem. This code is implemented as a module of a computation system used for reactor core history calculations. Given power density data, the geometric description in (RZ), and basic heat removal conditions and thermal properties, the coolant properties, flow conditions, and temperature distributions in the pebble fuel elements are predicted. The calculation is oriented to the continuous fueling, steady state condition with consideration of the effect of the high energy neutron flux exposure and temperature history on the thermal conductivity. The coolant flow conditions are calculated for the same geometry as used in the neutronics calculation, power density and fluence data being used directly, and temperature results are made available for subsequent use.

  16. Thermal fatigue durability for advanced propulsion materials

    Science.gov (United States)

    Halford, Gary R.

    1989-01-01

    A review is presented of thermal and thermomechanical fatigue (TMF) crack initiation life prediction and cyclic constitutive modeling efforts sponsored recently by the NASA Lewis Research Center in support of advanced aeronautical propulsion research. A brief description is provided of the more significant material durability models that were created to describe TMF fatigue resistance of both isotropic and anisotropic superalloys, with and without oxidation resistant coatings. The two most significant crack initiation models are the cyclic damage accumulation model and the total strain version of strainrange partitioning. Unified viscoplastic cyclic constitutive models are also described. A troika of industry, university, and government research organizations contributed to the generation of these analytic models. Based upon current capabilities and established requirements, an attempt is made to project which TMF research activities most likely will impact future generation propulsion systems.

  17. Best estimate methods in thermal hydraulic safety analysis - Seminar Summary and Conclusions

    International Nuclear Information System (INIS)

    In the spring of 1992 an OECD Specialist Meeting on Transient Two-Phase Flow ('Current issue in system thermal-hydraulics', April 6-8, 1992) was held in Aix-en-Provence, France. Next, there was an OECD Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements (Annapolis, USA, November 5-8, 1996). The issues raised during the meetings indicated strongly a need for well established best estimate methodology for the use in plant safety analysis. A number of other CSNI activities addressed status of codes validation and related issues of codes uncertainties. The CSNI Seminar on Best Estimate Methods in Thermal Hydraulic Safety Analysis, held on 29 June - 1 July 1998 in Ankara, Turkey, was sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organized in collaboration with Turkish Atomic Energy Authority (TAEK). The scope of this Seminar is limited to safety analysis needed in support of licensing process. Therefore, the workshop did not specifically address issues related to code development and physical models. The objectives of the meeting were: - to exchange information on the Member countries' methodologies used and/or required in the licensing process, and - to discuss the licensing issues associated with uncertainties and evaluation of T/H safety margins in conservative and BE approaches. - to provide information to Turkish hosts on the use of Best Estimate methods in support of licensing. The Seminar was structured into the following sessions: 1. Keynote presentations; 2. Overview of past and current CSNI Thermal-Hydraulics activities; 3. Best Estimate Methodologies and associated uncertainties (2 Sessions); 4. Selected Issues of Thermal-Hydraulics Safety Analyses; 5. Final Discussion. The studies of beyond design basis accident (BDBA) and severe accident (SA) scenarios are very important to evaluate a level of 'enhanced' plant safety. As it is difficult to calculate exactly the

  18. Development of subchannel analysis code MATRA-LMR for KALIMER subassembly thermal-hydraulics

    International Nuclear Information System (INIS)

    In the sodium cooled liquid metal reactors, the design limit are imposed on the maximum temperatures of claddings and fuel pins. Thus an accurate prediction of core coolant/fuel temperature distribution is essential to the LMR core thermal-hydraulic design. The detailed subchannel thermal-hydraulic analysis code MATRA-LMR (Multichannel Analyzer for Steady States and Transients in Rod Arrays for Liquid Metal Reactors) is being developed for KALIMER core design and analysis, based on COBRA-IV-i and MATRA. The major modifications and improvements implemented into MATRA-LMR are as follows: a) nonuniform axial noding capability, b) sodium properties calculation subprogram, c) sodium coolant heat transfer correlations, and d) most recent pressure drop correlations, such as Novendstern, Chiu-Rohsenow-Todreas and Cheng-Todreas. To assess the development status of this code, the benchmark calculations were performed with the ORNL 19 pin tests and EBR-II seven-assembly SLTHEN calculation results. The calculation results of MATRA-LMR for ORNL 19-pin assembly tests and EBR-II 91-pin experiments were compared to the measurements, and to SABRE4 and SLTHEN code calculation results, respectively. In this comparison, the differences are found among the three codes because of the pressure drop and the thermal mixing modellings. Finally, the major technical results of the conceptual design for the KALIMER 98.03 core have been compared with the calculations of MATRA-LMR, SABRE4 and SLTHEN codes. (author)

  19. Thermal-hydraulic performance of novel louvered fin using flat tube cross-flow heat exchanger

    Institute of Scientific and Technical Information of China (English)

    Junqi DONG; Jiangping CHEN; Zhijiu CHEN

    2008-01-01

    Experimental studies were conducted to investigate the air-side heat transfer and pressure drop characteristics of a novel louvered fins and flat tube heat exchangers. A series of tests were conducted for 9 heat exchangers with different fin space and fin length, at a constant tube-side water flow rate of 2.8 m/h. The air side thermal performance data were analyzed using the effectiveness-NTU method. Results were presented as plot of Colburn j factor and friction factor f against the Reynolds number in the range of 500-6500. The characteristics of the heat transfer and pressure drop of different fin space and fin length were analyzed and compared. In addition, the curves of the heat transfer coefficients vs. pumping power per unit heat transfer area were plotted. Finally, the area optimization factor was used to evaluate the thermal hydraulic performance of the louvered fins with differential geometries. The results showed that the j and ffactors increase with the decrease of the fin space and fin length, and the fin space has more obvious effect on the thermal hydraulic characteristics of the novel louvered fins.

  20. Thermal-hydraulic analysis on Ex-Vessel fuel Storage Tank of MONJU at severe accident

    International Nuclear Information System (INIS)

    In this paper, results of a thermal-hydraulic analysis on the Ex-Vessel fuel Storage Tank (EVST) of the fast breeder reactor MONJU at severe accident is described. Safety evaluations on this facility have ever been performed by using a one-dimensional flow network code. However, validation on a model of this code has been needed, because EVST has plenums and asymmetry equipment. Therefore we performed a CFD analysis under a condition of station blackout (SBO) in order to clarify the circulation flow rate and multidimensionality of the EVST. As a result, the following points were confirmed: 1) Circulation flow rate is maintained half of a flow rate at the rated operation condition at the minimum. 2) Thermal stratification arises in the lower plenum at SBO. 3) Circumferential distribution of flow rate at the lower plenum is made uniform at the inlet of the rotating rack. 4) Thermal-hydraulic behavior in the rotating rack is almost one-dimensional. (author)