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Sample records for thermal hydraulic performance

  1. Thermal-Hydraulic Performance of Scrubbing Nozzle Used for CFVS

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyun Chul; Lee, Doo Yong; Jung, Woo Young; Lee, Jong Chan; Kim, Gyu Tae [FNC TECH, Yongin (Korea, Republic of)

    2016-05-15

    A Containment Filtered Venting System (CFVS) is the most interested device to mitigate a threat against containment integrity under the severe accident of nuclear power plant by venting with the filtration of the fission products. FNC technology and partners have been developed the self-priming scrubbing nozzle used for the CFVS which is based on the venturi effect. The thermal-hydraulic performances such as passive scrubbing water suction as well as pressure drop across the nozzle have been tested under various thermal-hydraulic conditions. The two types of test section have been built for testing the thermal-hydraulic performance of the self-priming scrubbing nozzle. Through the visualization loop, the liquid suction performance through the slit, pressure drop across the nozzle are measured. The passive water suction flow through the suction slit at the throat is important parameter to define the scrubbing performance of the self-priming scrubbing nozzle. The water suction flow is increased with the increase of the overhead water level at the same inlet gas flow. It is not so much changed with the change of inlet gas flow at the overhead water level.

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

    International Nuclear Information System (INIS)

    1988-07-01

    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

  3. Development of fuel performance and thermal hydraulic technology

    International Nuclear Information System (INIS)

    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)

  4. 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. Copyright © 2015 Elsevier Ltd. All rights reserved.

  5. COMMIX analysis of four constant flow thermal upramp experiments performed in a thermal hydraulic model of an advanced LMR

    International Nuclear Information System (INIS)

    Yarlagadda, B.S.

    1989-04-01

    The three-dimensional thermal hydraulics computer code COMMIX-1AR was used to analyze four constant flow thermal upramp experiments performed in the thermal hydraulic model of an advanced LMR. An objective of these analyses was the validation of COMMIX-1AR for buoyancy affected flows. The COMMIX calculated temperature histories of some thermocouples in the model were compared with the corresponding measured data. The conclusions of this work are presented. 3 refs., 5 figs

  6. Development of Mitsubishi high thermal performance grid 1 - CFD applicability for thermal hydraulic design

    International Nuclear Information System (INIS)

    Ikeda, K.; Hoshi, M.

    2001-01-01

    Mitsubishi applied the Computational Fluid Dynamics (CFD) evaluation method for designing of the new lower pressure loss and higher DNB performance grid spacer. Reduction of pressure loss of the grid has been estimated by CFD. Also, CFD has been developed as a design tool to predict the coolant mixing ability of vane structures, that is to compare the relative peak spot temperatures around fuel rods at the same heat flux condition. These evaluations have been reflected to the new grid spacer design. The prototype grid was manufactured and some flow tests were performed to examine the thermal hydraulic performance, which were predicted by CFD. The experimental data of pressure loss was in good agreement with CFD prediction. The CFD prediction of flow behaviors at downstream of the mixing vanes was verified by detail cross-flow measurements at rod gaps by the rod LDV system. It is concluded that the applicability of the CFD evaluation method for the thermal hydraulic design of the grid is confirmed. (authors)

  7. Optimal thermal-hydraulic performance for helium-cooled divertors

    International Nuclear Information System (INIS)

    Izenson, M.G.; Martin, J.L.

    1996-01-01

    Normal flow heat exchanger (NFHX) technology offers the potential for cooling divertor panels with reduced pressure drops (<0.5% Δp/p), reduced pumping power (<0.75% pumping/thermal power), and smaller duct sizes than conventional helium heat exchangers. Furthermore, the NFHX can easily be fabricated in the large sizes required for divertors in large tokamaks. Recent experimental and computational results from a program to develop NFHX technology for divertor coolings using porous metal heat transfer media are described. We have tested the thermal and flow characteristics of porous metals and identified the optimal heat transfer material for the divertor heat exchanger. Methods have been developed to create highly conductive thermal bonds between the porous material and a solid substrate. Computational fluid dynamics calculations of flow and heat transfer in the porous metal layer have shown the capability of high thermal effectiveness. An 18-kW NFHX, designed to meet specifications for the international Thermonuclear Experimental Reactor divertor, has been fabricated and tested for thermal and flow performance. Preliminary results confirm design and fabrication methods. 11 refs., 12 figs., 1 tab

  8. Methodology for thermal hydraulic conceptual design and performance analysis of KALIMER core

    International Nuclear Information System (INIS)

    Young-Gyun Kim; Won-Seok Kim; Young-Jin Kim; Chang-Kue Park

    2000-01-01

    This paper summarizes the methodology for thermal hydraulic conceptual design and performance analysis which is used for KALIMER core, especially the preliminary methodology for flow grouping and peak pin temperature calculation in detail. And the major technical results of the conceptual design for the KALIMER 98.03 core was shown and compared with those of KALIMER 97.07 design core. The KALIMER 98.03 design core is proved to be more optimized compared to the 97.07 design core. The number of flow groups are reduced from 16 to 11, and the equalized peak cladding midwall temperature from 654 deg. C to 628 deg. C. It was achieved from the nuclear and thermal hydraulic design optimization study, i.e. core power flattening and increase of radial blanket power fraction. Coolant flow distribution to the assemblies and core coolant/component temperatures should be determined in core thermal hydraulic analysis. Sodium flow is distributed to core assemblies with the overall goal of equalizing the peak cladding midwall temperatures for the peak temperature pin of each bundle, thus pin cladding damage accumulation and pin reliability. The flow grouping and the peak pin temperature calculation for the preliminary conceptual design is performed with the modules ORFCE-F60 and ORFCE-T60 respectively. The basic subchannel analysis will be performed with the SLTHEN code, and the detailed subchannel analysis will be done with the MATRA-LMR code which is under development for the K-Core system. This methodology was proved practical to KALIMER core thermal hydraulic design from the related benchmark calculation studies, and it is used to KALIMER core thermal hydraulic conceptual design. (author)

  9. Thermal-hydraulic performance of convective boiling jet array impingement

    International Nuclear Information System (INIS)

    Jenkins, R; De Brún, C; Kempers, R; Lupoi, R; Robinson, A J

    2016-01-01

    Jet impingement boiling is investigated with regard to heat transfer and pressure drop performance using a novel laser sintered 3D printed jet impingement manifold design. Water was the working fluid at atmospheric pressure with inlet subcooling of 7 o C. The convective boiling performance of the impinging jet system was investigated for a flat copper target surface for 2700≤Re≤5400. The results indicate that the heat transfer performance of the impinging jet is independent of Reynolds number for fully developed boiling. Also, the investigation of nozzle to plate spacing shows that low spacing delays the onset of nucleate boiling causing a superheat overshoot that is not observed with larger gaps. However, no sensitivity to the gap spacing was measured once boiling was fully developed. The assessment of the pressure drop performance showed that the design effectively transfers heat with low pumping power requirements. In particular, owing to the insensitivity of the heat transfer to flow rate during fully developed boiling, the coefficient of performance of jet impingement boiling in the fully developed boiling regime deteriorates with increased flow rate due to the increase in pumping power flux. (paper)

  10. 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

  11. Thermal-Hydraulic Analysis of a Once-Through Steam Generator Considering Performance Degradation

    Energy Technology Data Exchange (ETDEWEB)

    Han, Hun Sik; Kang, Han Ok; Yoon, Ju Hyeon; Kim, Young In; Song, Jae Seung; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Several countries have entered into a global race for the commercialization of SMRs, and considerable research and development have been implemented. Among the various reactor designs, many SMRs have adopted an integral type pressurized water reactor (PWR) to enhance the nuclear safety and system reliability. In the integral reactor design, a single reactor pressure vessel contains primary system components such as fuel and core, steam generators, pumps, and a pressurizer. For the component integration into a reactor vessel, it is important to design each component as small as possible. Thus, it is a common practice to employ a once-through steam generator in the integral reactor design due to its advantages in compactness. In general, gradual degradation in thermal-hydraulic performance of the steam generator occurs with time, and it changes slowly the operating point of the steam generator during plant lifetime. Numerical solutions are acquired to evaluate the thermal-hydraulic performance of the steam generator at various AUFs. The design results obtained show that the average tube length of the steam generator is augmented with the increase of design margin to compensate for the design uncertainties and heat transfer area reduction by plugging, fouling, etc. A helically coiled tube once-through steam generator with 30% design margin is considered for comparison of thermal-hydraulic performances according to the degradation rate.

  12. Thermal-Hydraulic Analysis of a Once-Through Steam Generator Considering Performance Degradation

    International Nuclear Information System (INIS)

    Han, Hun Sik; Kang, Han Ok; Yoon, Ju Hyeon; Kim, Young In; Song, Jae Seung; Kim, Keung Koo

    2016-01-01

    Several countries have entered into a global race for the commercialization of SMRs, and considerable research and development have been implemented. Among the various reactor designs, many SMRs have adopted an integral type pressurized water reactor (PWR) to enhance the nuclear safety and system reliability. In the integral reactor design, a single reactor pressure vessel contains primary system components such as fuel and core, steam generators, pumps, and a pressurizer. For the component integration into a reactor vessel, it is important to design each component as small as possible. Thus, it is a common practice to employ a once-through steam generator in the integral reactor design due to its advantages in compactness. In general, gradual degradation in thermal-hydraulic performance of the steam generator occurs with time, and it changes slowly the operating point of the steam generator during plant lifetime. Numerical solutions are acquired to evaluate the thermal-hydraulic performance of the steam generator at various AUFs. The design results obtained show that the average tube length of the steam generator is augmented with the increase of design margin to compensate for the design uncertainties and heat transfer area reduction by plugging, fouling, etc. A helically coiled tube once-through steam generator with 30% design margin is considered for comparison of thermal-hydraulic performances according to the degradation rate

  13. The analysis of thermal-hydraulic performances of nuclear ship reactor

    International Nuclear Information System (INIS)

    Wakabayashi, Shinshichi; Hamada, Masao

    1975-01-01

    Thermal-hydraulic performances in the core of nuclear ship reactor was analysed by thermal-hydraulic analyser codes, AMRTC and COBRA-11+DNBCAL. This reactor is of a pressurized water type and incorporates the steam generator within the reactor vessel with the rated power of 330 MWt, which is developed by Nuclear Ship Research Panel Seven (NSR-7) in The Shipbuilding Research Association of Japan. Fuel temperature distributions, coolant temperature distributions, void fractions in coolant and minimum burn out ratio etc. were calculated. Results are as follows; a) The maximum temperature of fuel center is 1,472 0 C that corresponds to 53% as small as the melting point (2,800 0 C). b) Subcooled boiling exists in the core and the maximum void fraction is less than 4%. c) The minimum burn out ratio is not less than the minimum allowable limit of 1.25. It was found from the results of analysis that this reactor was able to be operated wide margin with respect to thermal-hydraulic design limits at the rated power. (auth.)

  14. Comparative performance of the conjugate gradient and SOR [Successive Over Relaxation] methods for computational thermal hydraulics

    International Nuclear Information System (INIS)

    King, J.B.; Anghaie, S.; Domanus, H.M.

    1987-01-01

    Finite difference approximations to the continuity, momentum, and energy equations in thermal hydraulics codes result in a system of N by N equations for a problem having N field points. In a three dimensional problem, N increases as the problem becomes larger or more complex, and more rapidly as the computational mesh size is reduced. As a consequence, the execution time required to solve the problem increases, which may lead to placing limits on the problem resolution or accuracy. A conventinal method of solution of these systems of equations is the Successive Over Relaxation (SOR) technique. However, for a wide range of problems the execution time may be reduced by using a more efficient linear equation solver. One such method is the conjugate gradient method which was implemented in COMMIX-1B thermal hydraulics code. It was found that the execution time required to solve the resulting system of equations was reduced by a factor of about 2 for some problems. This paper summarizes the characteristics of these iterative solution procedures and compares their performance in modeling of a variety of reactor thermal hydraulic problems, using the COMMIX-1B computer code

  15. Influence of sinusoidal flow on the thermal and hydraulic performance of microchannel heat sink

    International Nuclear Information System (INIS)

    Om, N I; Gunnasegaran, P; Rajasegaran, S

    2013-01-01

    In this paper, the effect of sinusoidal flow on the thermal and hydraulic performance of microchannel heat sink (MCHS) is numerically investigated. This investigation covers Reynolds number in the range of 100 ≤ Re ≤ 1000 and pure water is used as a working fluid. The three-dimensional steady, laminar flow and heat transfer governing equations are solved using finite volume method (FVM). The water flow field and heat transfer performance inside the sinusoidal microchannels is simulated and the results are compared with the straight microchannels. The effect of using sinusoidal microchannels on temperature distribution, Nusselt number, friction factor and thermal resistance is presented in this paper. It is found that with same rectangular cross-section, sinusoidal microchannels have a better heat transfer performance compared to the straight microchannels.

  16. Thermal-Hydraulic Performance of a Corrugated Cooling Fin with Louvered Surfaces

    DEFF Research Database (Denmark)

    Sønderby, Simon Kaltoft; Hosseini, Seyed Mojtaba Mir; Rezaniakolaei, Alireza

    2017-01-01

    The main objective of the article is to investigate thermal-hydraulic performance of a corrugated cooling fin with louvered surfaces. The investigation is carried out using the fin geometry of one most commonly used liquid-to-air heat exchangers. The investigation was carried out by numerically...... simulating the airflow with louvered fin geometry. The simulation model was verified by comparing simulated j- and f-factors with the corresponding values of several experimental correlations. The j-factors deviated less than 10.7 % from two of the experimental correlations, whereas deviations ranging...

  17. KUGEL: a thermal, hydraulic, fuel performance, and gaseous fission product release code for pebble bed reactor core analysis

    International Nuclear Information System (INIS)

    Shamasundar, B.I.; Fehrenbach, M.E.

    1981-05-01

    The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations

  18. Neutronic and Thermal-hydraulic Modelling of High Performance Light Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Seppaelae, Malla [VTT Technical Research Centre of Finland, P.O.Box 1000, FI02044 VTT (Finland)

    2008-07-01

    High Performance Light Water Reactor (HPLWR), which is studied in EU project 'HPLWR2', uses water at supercritical pressures as coolant and moderator to achieve higher core outlet temperature and thus higher efficiency compared to present reactors. At VTT Technical Research Centre of Finland, functionality of the thermal-hydraulics in the coupled reactor dynamics code TRAB3D/ SMABRE was extended to supercritical pressures for the analyses of HPLWR. Input models for neutronics and thermal-hydraulics were made for TRAB3D/ SMABRE according to the latest HPLWR design. A preliminary analysis was performed in which the capability of SMABRE in the transition from supercritical pressures to subcritical pressures was demonstrated. Parameterized two-group cross sections for TRAB3D neutronics were received from Hungarian Academy of Sciences KFKI Atomic Energy Research Institute together with a subroutine for handling them. PSG, a new Monte Carlo transport code developed at VTT, was also used to generate two-group constants for HPLWR and comparisons were made with the KFKI cross sections and MCNP calculations. (author)

  19. Neutronic and Thermal-hydraulic Modelling of High Performance Light Water Reactor

    International Nuclear Information System (INIS)

    Seppaelae, Malla

    2008-01-01

    High Performance Light Water Reactor (HPLWR), which is studied in EU project 'HPLWR2', uses water at supercritical pressures as coolant and moderator to achieve higher core outlet temperature and thus higher efficiency compared to present reactors. At VTT Technical Research Centre of Finland, functionality of the thermal-hydraulics in the coupled reactor dynamics code TRAB3D/ SMABRE was extended to supercritical pressures for the analyses of HPLWR. Input models for neutronics and thermal-hydraulics were made for TRAB3D/ SMABRE according to the latest HPLWR design. A preliminary analysis was performed in which the capability of SMABRE in the transition from supercritical pressures to subcritical pressures was demonstrated. Parameterized two-group cross sections for TRAB3D neutronics were received from Hungarian Academy of Sciences KFKI Atomic Energy Research Institute together with a subroutine for handling them. PSG, a new Monte Carlo transport code developed at VTT, was also used to generate two-group constants for HPLWR and comparisons were made with the KFKI cross sections and MCNP calculations. (author)

  20. GCFR thermal-hydraulic experiments

    International Nuclear Information System (INIS)

    Schlueter, G.; Baxi, C.B.; Dalle Donne, M.; Gat, U.; Fenech, H.; Hanson, D.; Hudina, M.

    1980-01-01

    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

  1. Effect of tube bank inclination on the thermal hydraulic performance of air-cooled heat exchangers

    International Nuclear Information System (INIS)

    Monheit, M.; Freim, J.

    1987-01-01

    This paper compares the experimental thermal hydraulic performance of an inclined tube bundle of staggered high finned tubes with that of the same bank in cross flow. The data are shown on a relative basis with the conventional cross-flow arrangement used as the reference layout. The results, presented as plots of heat transfer coefficient and pressure drop per row versus face velocity, are compared with other published data for bare and extended-surface tube banks. Correction factors that account for the deviation from cross-flow performance are presented. Estimated savings in plot area are shown for typical cases. Some possible cost reduction opportunities in specific fields of application are also suggested

  2. Effect of fin-endwall fillet on thermal hydraulic performance of airfoil printed circuit heat exchanger

    International Nuclear Information System (INIS)

    Ma, Ting; Xin, Fei; Li, Lei; Xu, Xiang-yang; Chen, Yi-tung; Wang, Qiu-wang

    2015-01-01

    Printed circuit heat exchanger (PCHE) is recommended to be used for intermediate heat exchanger in Very High Temperature Reactor (VHTR). One of the key features is that it is manufactured by the photochemical etching in order to maintain the internal structure and metal properties. In this paper, a photochemical etching experiment is conducted to manufacture the airfoil PCHE plate. The result indicates that the airfoil fin is not an ideal airfoil profile, but has a fin-endwall fillet. For the purpose of simplifying the numerical model and saving computational time, a validated model with a single fluid is used to further study the effect of fin-endwall fillet on the thermal hydraulic performance of airfoil PCHE. It is found that the fin-endwall fillet can increase the heat transfer and pressure drop in the cases with the non-dimensional longitudinal pitch being 1.63. The effect of fin-endwall fillet on thermal hydraulic performance decreases with the increase of transverse pitch, but the longitudinal pitch has little effect when the non-dimensional longitudinal pitch is greater than 1.88. In the studied cases, the maximum difference of Nusselt number and friction factor between the two models with and without fin-endwall fillet is up to 6.7% and 6.4%. - Highlights: • Fillets are formed in the endwall of airfoil fins during the photochemical etching. • Two-fluid model can be replaced by single-fluid model to perform simulation. • Fin-endwall fillet can increase heat transfer and pressure drop at ζ_l = 1.63. • Effect of fin-endwall fillet decreases as transverse pitch increases at ζ_l = 1.63. • Longitudinal pitch has little effect at ζ_l ≥ 1.88.

  3. Thermal Hydraulic Performance in a Solar Air Heater Channel with Multi V-Type Perforated Baffles

    Directory of Open Access Journals (Sweden)

    Anil Kumar

    2016-07-01

    Full Text Available This article presents heat transfer and fluid flow characteristics in a solar air heater (SAH channel with multi V-type perforated baffles. The flow passage has an aspect ratio of 10. The relative baffle height, relative pitch, relative baffle hole position, flow attack angle, and baffle open area ratio are 0.6, 8.0, 0.42, 60°, and 12%, respectively. The Reynolds numbers considered in the study was in the range of 3000–10,000. The re-normalization group (RNG k-ε turbulence model has been used for numerical analysis, and the optimum relative baffle width has been investigated considering relative baffle widths of 1.0–7.0.The numerical results are in good agreement with the experimental data for the range considered in the study. Multi V-type perforated baffles are shown to have better thermal performance as compared to other baffle shapes in a rectangular passage. The overall thermal hydraulic performance shows the maximum value at the relative baffle width of 5.0.

  4. Numerical simulation of the thermal hydraulic performance of a plate pin fin heat sink

    International Nuclear Information System (INIS)

    Yuan Wuhan; Zhao Jiyun; Tso, C.P.; Wu Tianhua; Liu Wei; Ming Tingzhen

    2012-01-01

    The computational fluid dynamic software FLUENT is used in assessing the electronics cooling potential of a plate pin fin heat sink (PPFHS), including the conjugate effect. The simulation results are validated with reported experimental data. The simulation shows that pin height and air velocity have significant influences on the thermal hydraulic performances of PPFHS while the influences of in-line/staggered array and neighbor pin flow-directional center distance (NPFDCD) of the PPFHS are less notable. In applying the present design to the cooling of a desktop PC CPU at a heat flux of 2.20 W/cm 2 , the temperature can be kept at less than 358 K with an air velocity over 6.5 m/s. - Highlights: ► Pin height and air velocity significantly influence thermal performance of PPFHS. ► Less influence by in-line or staggered array. ► Less influence by neighbor pin flow-directional center distance. ► Design with >6.5 m/s air can cool to 2 flux.

  5. Numerical investigation on thermal-hydraulic performance of new printed circuit heat exchanger model

    International Nuclear Information System (INIS)

    Kim, Dong Eok; Kim, Moo Hwan; Cha, Jae Eun; Kim, Seong O.

    2008-01-01

    Three-dimensional numerical analysis was performed to investigate heat transfer and pressure drop characteristics of supercritical CO 2 flow in new Printed Circuit Heat Exchanger (PCHE) model using commercial CFD code, Fluent 6.3. First, numerical analysis for conventional zigzag channel PCHE model was performed and compared with previous experimental data. Maximum deviation of in-outlet temperature difference and pressure drop from experimental data is about 10%. A new PCHE model has been designed to optimize thermal-hydraulic performance of PCHE. The new PCHE model has several airfoil shape fins (NACA 0020 model), which are designed to streamlined shape. Simulation results showed that in the airfoil shape fin PCHE, total heat transfer rate per unit volume was almost same with zigzag channel PCHE and the pressure drop was reduced to one-twentieth of that in zigzag channel PCHE. In airfoil shape fin PCHE model, the enhancement of heat transfer area and the uniform flow configuration contributed to obtain the same heat transfer performance with zigzag channel PCHE model. And the reduction of pressure drop in airfoil shape fin PCHE model was caused by suppressing generation of separated flow owing to streamlined shape of airfoil fins

  6. Numerical investigation of thermal-hydraulic performance of channel with protrusions by turbulent cross flow jet

    Science.gov (United States)

    Sahu, M. K.; Pandey, K. M.; Chatterjee, S.

    2018-05-01

    In this two dimensional numerical investigation, small rectangular channel with right angled triangular protrusions in the bottom wall of test section is considered. A slot nozzle is placed at the middle of top wall of channel which impinges air normal to the protruded surface. A duct flow and nozzle flow combined to form cross flow which is investigated for heat transfer enhancement of protruded channel. The governing equations for continuity, momentum, energy along with SST k-ω turbulence model are solved with finite volume based Computational fluid dynamics code ANSYS FLUENT 14.0. The range of duct Reynolds number considered for this analysis is 8357 to 51760. The ratios of pitch of protrusion to height of duct considered are 0.5, 0.64 and 0.82. The ratios of height of protrusion to height of duct considered are 0.14, 0.23 and 0.29. The effect of duct Reynolds number, pitch and height of protrusion on thermal-hydraulic performance is studied under cross flow condition. It is found that heat transfer rate is more at relatively larger pitch and small pressure drop is found in case of low height of protrusion.

  7. Thermal-hydraulic unreliability of passive systems

    International Nuclear Information System (INIS)

    Tzanos, C.P.; Saltos, N.T.

    1995-01-01

    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

  8. Steam generator thermal-hydraulics

    International Nuclear Information System (INIS)

    Inch, W.W.; Scott, D.A.; Carver, M.B.

    1980-01-01

    This paper discusses a code for detailed numerical modelling of steam generator thermal-hydraulics, and describes related experimental programs designed to promote in-depth understanding of three-dimensional two-phase flow. (auth)

  9. A numerical study on improving the thermal hydraulic performance of printed circuit heat exchanger using the supercritical carbon dioxide

    Energy Technology Data Exchange (ETDEWEB)

    Park, Bo Guen; Kim, Dae Hyun; Chung, Jin Taek [Dept. of Mechanical Engineering, Korea University, Seoul (Korea, Republic of)

    2015-10-15

    The objective of this study is to propose a new channel shape that improves thermal-hydraulic performance. The existing Zigzag channel has high pressure loss due to flow separation and reverse flow. To improve this disadvantage, partial straight channel is inserted into bended points. Also, the effects of straight channel's length change on heat transfer and pressure loss are analyzed. Thermal-hydraulic performance of the new shape and existing Zigzag channel are quantitatively compared in terms of Goodness Factor. Mass flow rate was changed from 1.41 x 10{sup -4} kg/s to 2.48 x 10{sup -4} kg/s . The average volume goodness factor of 1mm straight channel shape was increased by 25% compared to the Zigzag channel.

  10. Feasibility study on thermal-hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Ohnuki, A.; Kureta, M.; Liu, W.; Tamai, H.; Akimoto, H.

    2004-01-01

    Research and development project for investigating thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in 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 light-water reactor 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 issues for the RMWR because of the tight-lattice configuration. The project has mainly consisted of a large-scale thermal-hydraulic test and development of analytical methods named modeling engineering. In the large-scale test, 37-rod bundle experiments can be performed. Steady-state critical power experiments have been achieved in the test facility and the experimental data reveal the feasibility of RMWR

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

    International Nuclear Information System (INIS)

    Conboy, Thomas; Hejzlar, Pavel

    2006-01-01

    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

  12. The examination of the effects of the flow baffle on the thermal hydraulic performance of IHX shell side

    International Nuclear Information System (INIS)

    Kim, W. K.; Sim, Y. S.; Kim, S. O.; Baek, B. J.

    2002-01-01

    The effects of the flow baffle on the thermal hydraulic performance of IHX shell side has been examined using ASTEEPL, 2DHX code for the variation of baffle distance, baffle cut and baffle flow hole design data. When KALIMER design data were selected as a base for the study, a shell baffle structures does not influence on a total heat transfer rate and it is better to reduce baffle distance rather than baffle flow hole size from the view of pressure loss in the IHX. Radial thermal imbalance between tubes was reduced to 6 .deg. C when the number of baffle is beyond 13. And thermal imbalance decreased also when flow hole size decreases. If flow hole clogs perfectly, thermal imbalance can be reduced up to 2 .deg. C

  13. Thermal-hydraulic performance of the finned surface of a compact heat exchanger

    International Nuclear Information System (INIS)

    Errasti Cabrera, Michel

    2015-01-01

    In this work the thermal-hydraulic behavior of the finned surface of a compact heat exchanger is obtained in tube-fin configuration corrugated (wavy). Through numerical simulation are determined average values ​​of intensification of heat transfer and pressure loss in the inter-channel finned. The objective is to characterize the surface to use as a reference, to make comparisons with other heat exchange surfaces enhanced using traditional techniques combined with more current, such as vortex generators. The study is conducted in laminar flow, with Reynolds numbers below 1000. In the working model compact exchanger tubes and corrugated fins (wavy) heat is described, and the results of the coefficient of overall heat transfer and the pressure drop are explained from the local characteristics of the velocity field and temperature inside the heat exchanger. (Full text)

  14. Coupled neutronics/thermal-hydraulics analysis of a high-performance light-water reactor fuel assembly

    International Nuclear Information System (INIS)

    Waata, C.L.

    2006-07-01

    The use of water at supercritical pressure as coolant and moderator introduces a challenge in the design of a High-Performance Light-Water Reactor (HPLWR) fuel assembly. At supercritical pressure condition (P=25 MPa), the thermal-hydraulics behaviour of water differs strongly from that at sub-critical pressure due to a rapid variation of the thermal-physical properties across the pseudo-critical line. Due of the strong link between the water (moderation) and the neutron spectrum and subsequently the power distribution, a coupling of neutronics and thermal-hydraulics has become a necessity for reactor concepts operating at supercritical pressure condition. The effect of neutron moderation on the local parameters of thermal-hydraulics and vice-verse in a fuel assembly has to be considered for an accurate design analysis. In this study, the Monte Carlo N-Particle code (MCNP) and the sub-channel code STAFAS (Sub-channel Thermal-hydraulics Analysis of a Fuel Assembly under Supercritical conditions) have been coupled for the design analysis of a fuel assembly with supercritical water as coolant and moderator. Both codes are well known for complex geometry modelling. The MCNP code is used for neutronics analyses and for the prediction of power profiles of individual fuel rods. The sub-channel code STAFAS for the thermal-hydraulics analyses takes into account the coolant properties beyond the critical point as well as separate moderator channels. The coupling procedure is realized automatically. MCNP calculates the power distribution in each fuel rod, which is then transferred into STAFAS to obtain the corresponding thermal-hydraulic conditions in each sub-channel. The new thermal-hydraulic conditions are used to generate a new input deck for the next MCNP calculation. This procedure is repeated until a converged state is achieved. The coupled code system was tested on a proposed fuel assembly design of a HPLWR. An under-relaxation was introduced to achieve convergence

  15. Thermal hydraulics and mechanics core design programs

    International Nuclear Information System (INIS)

    Heinecke, J.

    1992-10-01

    The report documents the work performed within the Research and Development Task T hermal 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

  16. Liquid metal thermal-hydraulics

    International Nuclear Information System (INIS)

    Kottowski-Duemenil, H.M.

    1994-01-01

    This textbook is a report of the 26 years activity of the Liquid Metal Boiling Working Group (LMBWG). It summarizes the state of the art of liquid metal thermo-hydraulics achieved through the collaboration of scientists concerned with the development of the Fast Breeder Reactor. The first chapter entitled ''Liquid Metal Boiling Behaviour'', presents the background and boiling mechanisms. This section gives the reader a brief but thorough survey on the superheat phenomena in liquid metals. The second chapter of the text, ''A Review of Single and Two-Phase Flow Pressure Drop Studies and Application to Flow Stability Analysis of Boiling Liquid Metal Systems'' summarizes the difficulty of pressure drop simulation of boiling sodium in core bundles. The third chapter ''Liquid Metal Dry-Out Data for Flow in Tubes and Bundles'' describes the conditions of critical heat flux which limits the coolability of the reactor core. The fourth chapter dealing with the LMFBR specific topic of ''Natural Convection Cooling of Liquid Metal Systems''. This chapter gives a review of both plant experiments and out-of-pile experiments and shows the advances in the development of computing power over the past decade of mathematical modelling ''Subassembly Blockages Suties'' are discussed in chapter five. Chapter six is entitled ''A Review of the Methods and Codes Available for the Calculation on Thermal-Hydraulics in Rod-Cluster and other Geometries, Steady state and Transient Boiling Flow Regimes, and the Validation achieves''. Codes available for the calculation of thermal-hydraulics in rod-clusters and other geometries are reviewed. Chapter seven, ''Comparative Studies of Thermohydraulic Computer Code Simulations of Sodium Boiling under Loss of Flow Conditions'', represents one of the key activities of the LMBWG. Several benchmark exercises were performed with the aim of transient sodium boiling simulation in single channels and bundle blockages under steady state conditions and loss of

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

    International Nuclear Information System (INIS)

    Akira Ohnuki; Kazuyuki Takase; Masatoshi Kureta; Hiroyuki Yoshida; Hidesada Tamai; Wei Liu; Toru Nakatsuka; Hajime Akimoto

    2005-01-01

    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)

  18. First vapor explosion calculations performed with MC3D thermal-hydraulic code

    Energy Technology Data Exchange (ETDEWEB)

    Brayer, C.; Berthoud, G. [CEA Centre d`Etudes de Grenoble, 38 (France). Direction des Reacteurs Nucleaires

    1998-01-01

    This paper presents the first calculations performed with the `explosion` module of the multiphase computer code MC3D, which is devoted to the fine fragmentation and explosion phase of a fuel coolant interaction. A complete description of the physical laws included in this module is given. The fragmentation models, taking into account two fragmentation mechanisms, a thermal one and an hydrodynamic one, are also developed here. Results to some calculations to test the numerical behavior of MC3D and to test the explosion models in 1D or 2D are also presented. (author)

  19. Thermal-hydraulic performance of a water-cooled tungsten-rod target for a spallation neutron source

    International Nuclear Information System (INIS)

    Poston, D.I.

    1997-08-01

    A thermal-hydraulic (T-H) analysis is conducted to determine the feasibility and limitations of a water-cooled tungsten-rod target at powers of 1 MW and above. The target evaluated has a 10-cm x 10-cm cross section perpendicular to the beam axis, which is typical of an experimental spallation neutron source - both for a short-pulse spallation source and long-pulse spallation source. This report describes the T-H model and assumptions that are used to evaluate the target. A 1-MW baseline target is examined, and the results indicate that this target should easily handle the T-H requirements. The possibility of operating at powers >1 MW is also examined. The T-H design is limited by the condition that the coolant does not boil (actual limits are on surface subcooling and wall heat flux); material temperature limits are not approached. Three possible methods of enhancing the target power capability are presented: reducing peak power density, altering pin dimensions, and improving coolant conditions (pressure and temperature). Based on simple calculations, it appears that this target concept should have little trouble reaching the 2-MW range (from a purely T-H standpoint), and possibly much higher powers. However, one must keep in mind that these conclusions are based solely on thermal-hydraulics. It is possible, and perhaps likely, that target performance could be limited by structural issues at higher powers, particularly for a short-pulse spallation source because of thermal shock issues

  20. A HISTORICAL PERSPECTIVE OF NUCLEAR THERMAL HYDRAULICS

    Energy Technology Data Exchange (ETDEWEB)

    D’Auria, F; Rohatgi, Upendra S.

    2017-01-12

    The nuclear thermal-hydraulics discipline was developed following the needs for nuclear power plants (NPPs) and, to a more limited extent, research reactors (RR) design and safety. As in all other fields where analytical methods are involved, nuclear thermal-hydraulics took benefit of the development of computers. Thermodynamics, rather than fluid dynamics, is at the basis of the development of nuclear thermal-hydraulics together with the experiments in complex two-phase situations, namely, geometry, high thermal density, and pressure.

  1. Experimental study on the thermal hydraulic performance of plate-fin heat exchangers for cryogenic applications

    Science.gov (United States)

    Jiang, Qingfeng; Zhuang, Ming; Zhang, Qiyong; Zhu, Zhigang; Geng, Maofei; Sheng, Linhai; Zhu, Ping

    2018-04-01

    Efficient and compact plate-fin heat exchangers are critical for large-scale helium liquefaction/refrigeration systems as they constitute major part in the cold box. This study experimentally explores the heat transfer and pressure drop behaviors of helium gas at low temperature in four types of plate-fin channels, namely offset-strip and perforated fins, with different geometrical parameters. A series of cryogenic experiments at approximately liquid nitrogen temperature are carried out to measure the Colburn j factors and Fanning friction f factors with a wide range of Reynolds number. Besides, to reveal the performance variations under different operating temperatures, comparative experiments respectively conducted at room temperature and liquid nitrogen temperature are implemented. The results show that in comparison with the performance data at room temperature, most of j factors are relatively smaller perhaps because the lower aluminum thermal conductivity and higher Prandtl Number at low temperature. Meanwhile, the f factors corresponding to cryogenic conditions exhibit slightly larger even though the core pressure drops show considerable reductions. In contrast to the calculated results from the frequently-used performance curves (Chen and Shen, 1993), the Root Mean Squared Errors of j and f values are correlated within 8.38% and 6.97% for one perforated fin core, 41.29% and 34.97% for three OSF cores, respectively. For OSFs, further comparisons with the previous empirical correlations from literatures are conducted to verify the accuracy of each correlation. Generally, most of the calculated results predict acceptably within the deviations of ±25% for the j factors, while the predicted results express relatively large deviations for the f factors. Therefore, it may be revealed that most of the existing correlations were not able to accurately predict the experimental data in consideration of the performance differences under realistic cryogenic operating

  2. Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles: I-Master Plan and Executive Summary

    Science.gov (United States)

    Ohnuki, Akira; Kureta, Masatoshi; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, Wei; Misawa, Takeharu; Takase, Kazuyuki; Akimoto, Hajime

    R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Innovative Water Reactor for Flexible Fuel Cycle has been progressed at Japan Atomic Energy Agency in collaboration with power utilities, reactor vendors and universities since 2002. In this series-study, we will summarize the R&D achievements using large-scale test facility (37-rod bundle with full-height and full-pressure), model experiments and advanced numerical simulation technology. This first paper described the master plan for the development of design technology and showed an executive summary for this project up to FY2005. The thermal-hydraulic characteristics in the tight-lattice configuration were investigated and the feasibility was confirmed based on the experiments. We have developed the design technology including 3-D numerical simulation one to evaluate the effects of geometry/scale on the thermal-hydraulic behaviors.

  3. Reactor Thermal Hydraulic Numerical Calculation And Modeling

    International Nuclear Information System (INIS)

    Duong Ngoc Hai; Dang The Ba

    2008-01-01

    In the paper the results of analysis of thermal hydraulic state models using the numerical codes such as COOLOD, EUREKA and RELAP5 for simulation of the reactor thermal hydraulic states are presented. The calculations, analyses of reactor thermal hydraulic state and safety were implemented using different codes. The received numerical results, which were compared each to other, to experiment measurement of Dalat (Vietnam) research reactor and published results, show their appropriateness and capacity for analyses of different appropriate cases. (author)

  4. Thermal hydraulic model validation for HOR mixed core fuel management

    International Nuclear Information System (INIS)

    Gibcus, H.P.M.; Vries, J.W. de; Leege, P.F.A. de

    1997-01-01

    A thermal-hydraulic core management model has been developed for the Hoger Onderwijsreactor (HOR), a 2 MW pool-type university research reactor. The model was adopted for safety analysis purposes in the framework of HEU/LEU core conversion studies. It is applied in the thermal-hydraulic computer code SHORT (Steady-state HOR Thermal-hydraulics) which is presently in use in designing core configurations and for in-core fuel management. An elaborate measurement program was performed for establishing the core hydraulic characteristics for a variety of conditions. The hydraulic data were obtained with a dummy fuel element with special equipment allowing a.o. direct measurement of the true core flow rate. Using these data the thermal-hydraulic model was validated experimentally. The model, experimental tests, and model validation are discussed. (author)

  5. CFD Analysis on the Thermal Hydraulic Performance of an SAH Duct with Multi V-Shape Roughened Ribs

    Directory of Open Access Journals (Sweden)

    Anil Kumar

    2016-05-01

    Full Text Available This study presents the heat transfer and fluid flow characteristics in a rib-roughened SAH (solar air heater channel. The artificial roughness of the rectangular channel was in the form of a thin circular wire in discrete multi V-pattern rib geometries. The effect of this geometry on heat transfer, fluid flow, and performance augmentation was investigated using the CFD (computational fluid dynamics. The roughness parameters were a relative discrete distance of 0.69, a relative rib height of 0.043, a relative rib pitch of 10, a relative rib width of 6.0, and a flow-attack-angle of 60°. The discrete width ratios and Reynolds numbers ranged from 0.5 to 2.0 and from 2000 to 20,000, respectively. The CFD results using the renormalization k-epsilon model were in good agreement with the empirical relationship. This model was used to investigate the heat transfer and fluid flow characteristics in the multi V-pattern rib roughened SAH channel. The thermo-hydraulic performance was found to be the best for the discrete width ratio of 1.0. A discrete multi V-pattern rib combined with dimple staggered ribs also had better overall thermal performance compared to other rib shapes.

  6. HANARO thermal hydraulic accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chul; Kim, Heon Il; Lee, Bo Yook; Lee, Sang Yong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

    For the safety assessment of HANARO, accident analyses for the anticipated operational transients, accident scenarios and limiting accident scenarios were conducted. To do this, the commercial nuclear reactor system code. RELAP5/MOD2 was modified to RELAP5/KMRR; the thermal hydraulic correlations and the heat exchanger model was changed to incorporate HANARO characteristics. This report summarizes the RELAP/KMRR calculation results and the subchannel analyses results based on the RELAP/KMRR results. During the calculation, major concern was placed on the integrity of the fuel. For all the scenarios, the important accident analysis parameters, i.e., fuel centerline temperatures and the minimum critical heat flux ratio(MCHFR), satisfied safe design limits. It was verified, therefore, that the HANARO was safely designed. 21 tabs., 89 figs., 39 refs. (Author) .new.

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

    International Nuclear Information System (INIS)

    Song, C. H.; Baek, W. P.; Chung, M. K.

    2007-06-01

    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

  8. The Thermal-hydraulic Performance Test Report for the Non-instrumented Irradiation Test Rig of Annular Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Lee, Kang Hee; Shin, Chang Hwan

    2008-09-15

    This report presents the results of pressure drop test, vibration test and endurance test for the non-instrumented rig using the irradiation test in HANARO of the double cooled annular fuel which were designed and fabricated by KAERI. From the out-pile thermal hydraulic tests, corresponding to the pressure drop of 200 kPa is measured to be about 9.72 kg/sec. Vibration frequency for the non-instrumented rig ranges from 5.0 to 10.7 kg/s. RMS(Root Mean Square) displacement for non-instrumented rig is less than 11.73 m, and the maximum displacement is less than 54.87m. The flow rate for endurance test were 10.5 kg/s, which was 110% of 9.72 kg/s. And the endurance test was carried out for 3 days. The test results found not to the wear and satisfied to the limits of pressure drop, flow rate, vibration and wear in the non-instrumented rig. This test was performed at the FIVPET facility.

  9. Simulation of the fuel rod thermal hydraulic performance during the blow down phase in a PWR

    International Nuclear Information System (INIS)

    Gadelha, J.A.M.

    1982-10-01

    A digital computer code to predict the fuel rod thermalhydraulic performance during a postulated loss-of-coolant accident (LOCA) in the primary circuit of a PWR nuclear power plant is developed. The fuel rod corresponds to that in an average channel in the core. Only the blowdown phase is considered during the accident. The conservation equations of mass, momentum, and energy, and the heat conduction equation are solved to determine the fuel rod conditions during the accident. Finite differences are applied as a numerical method in the solution of the equations modelling the rod and coolant conditions. (Author) [pt

  10. Coupled fuel performance and thermal-hydraulics simulation with BISON and RELAP-7 at accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Martineau, R.C., E-mail: Richard.Martineau@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States)

    2015-07-01

    'Full text:' RELAP-7 is expected to be the next in the RELAP nuclear reactor safety/systems analysis application series developed at the Idaho National Laboratory (INL). The development of RELAP-7 began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of Department of Energy's (DOE) Light Water Reactor Sustainability (LWRS) Program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support modern nuclear power safety analysis. RELAP-7 is built using the INL's modern scientific software development framework, MOOSE (Multi-physics Object Oriented Simulation Environment). MOOSE provides improved implicit numerical schemes, including higher-order integration in both space and time, and yielding converged second-order accuracy for RELAP-7. The code structure is based on multiple physical component models such as pipes, junctions, pumps, etc. This component-based software architecture allows RELAP-7 to quickly adopt different physical models for different applications. One of the main advantages of building RELAP-7 on the MOOSE framework is that tight coupling with other MOOSE-based applications solving physics not present in RELAP-7 requires little to no additional lines of code. For example, the RELAP-7 core channel component is based upon a one-dimensional flow channel and a three-zone two-dimensional heat structure designed to represent fuel, gap, and cladding conjugate heat transfer with the coolant. However, the RELAP-7 application does not carry the fuels performance physics to analyze irradiated fuel, especially for accident scenarios. Here, we demonstrate the tightly coupled capability of the BISON nuclear fuels performance application with RELAP-7 for the station black out (SBO) accident scenario (Fukushima type event) and

  11. Thermal Hydraulic Design of PWT Accelerating Structures

    CERN Document Server

    Yu, David; Chen Ping; Lundquist, Martin; Luo, Yan

    2005-01-01

    Microwave power losses on the surfaces of accelerating structures will transform to heat which will deform the structures if it is not removed in time. Thermal hydraulic design of the disk and cooling rods of a Plane Wave Transformer (PWT) structure is presented. Experiments to measure the hydraulic (pressure vs flow rate) and cooling (heat removed vs flow rate) properties of the PWT disk are performed, and results compared with simulations using Mathcad models and the COSMOSM code. Both experimental and simulation results showed that the heat deposited on the structure could be removed effectively using specially designed water-cooling circuits and the temperature of the structure could be controlled within the range required.

  12. Thermal hydraulic feasibility assessment of the spent nuclear fuel project

    International Nuclear Information System (INIS)

    Heard, F.J.

    1996-01-01

    A series of analyses have been completed investigating the thermal-hydraulic performance and feasibility of the Spent Nuclear Fuel Project (SNFP) Integrated Process Strategy (IPS). The goal was to develop a series of thermal-hydraulic models that could respond to all process and safety related issues that may arise pertaining to the SNFP, as well as provide a basis for validation of the results. Results show that there is a reasonable envelope for process conditions and requirements that are thermally and hydraulically acceptable

  13. Experimental thermal hydraulics in support of FBR

    International Nuclear Information System (INIS)

    Padmakumar, G.; Anand Babu, C.; Kalyanasundaram, P.; Vaidyanathan, G.

    2009-01-01

    The thermal hydraulic design plays a crucial role for the safe and economical deployment of Liquid Metal Cooled Fast Breeder Reactor (LMFBR). Robust experimental programmes are required in support of LMFBR thermal hydraulics design. The philosophy of testing has been to construct small scale models to understand the physical behaviour and to build larger scale models to optimize the component design. The experiments are conducted either in sodium or using a simulant like water/air. The paper gives a brief account of the various thermal hydraulic experiments carried out in support of the design of Prototype Fast Breeder Reactor (PFBR). (author)

  14. Application of RELAP5/MOD3.3 to Calculate Thermal Hydraulic Behavior of the Pressurizer Safety Valve Performance Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyun; Kim, Young Ae; Oh, Seung Jong; Park, Jong Woon [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2007-10-15

    The increase of the acceptance tolerance of Pressurizer Safety Valve (PSV) test is vital for the safe operation of nuclear power plants because the frequent tests may make the valves decrepit and become a cause of leak. Recently, Korea Hydro and Nuclear Power Company (KHNP) is building a PSV performance test facility to provide the technical background data for the relaxation of the acceptance tolerance of PSV including the valve pop-up characteristics and the loop seal dynamics (if the plant has the loop seal in the upstream of PSV). The discharge piping and supports must be designed to withstand severe transient hydrodynamic loads when the safety valve actuates. The evaluation of hydrodynamic loads is a two-step process: first the thermal hydraulic behavior in the piping must be defined, and then the hydrodynamic loads are calculated from the thermal hydraulic parameters such as pressure and mass flow. The hydrodynamic loads are used as input to the structural analysis.

  15. Experimental and numerical study on thermal-hydraulic performance of wing-shaped-tubes-bundle equipped with winglet vortex generators

    Science.gov (United States)

    Abdelatief, Mohamed A.; Sayed Ahmed, Sayed Ahmed E.; Mesalhy, Osama M.

    2018-03-01

    The present work evaluates, experimentally and numerically, by the aid of commercial code FLUENT 6.3.26, the effects of relative locations (ΔX or ΔY), heights (hw), and span-angle (θ) of winglet-vortex-generators (WVGs) on thermal-hydraulic performance enhancement for down-stream and/or up-stream wing-shaped-tubes bundle heat exchangers for air Re ranging from 1.85 × 103 to 9.7 × 103 while water Re = 5 × 102. hw is set as 5 mm, 7.5 mm and 10 mm. For tube down-stream, θ is set as 0° (Base-line-case) and from 5° to 45° clockwise common-flow up (CFUp) and counterclockwise common-flow down (CFDn) while for tube up-stream it is set as -5°, -10° and -15° CFUp. Results show that the increase of θ counterclockwise-(CFDn) or clockwise-(CFUp) leads to increase the values of Nu number. Using WVGs with (+5 ° ≤ θ ≤ +45°) results in increasing Nu number by about from 34 to 48% comparing with that of base-line-case. The lowest values of drag coefficient ( f) for tube down-stream are obtained at +5° CFDn and -15° CFUp with respect to the base-line case. For tube up-stream, Nu number increases by increasing θ from 0° to -5° and the values of Nu number for θ varying from -5° to -15° have no significant changes. ( f) increases with hw and has negligible effect on ha. Furthermore, optimization analyses of θ and longitudinal fin (LF) are utilized, in four cases, for finding the optimum combination and maximum efficiency. The highest values of heat transfer parameters such as effectiveness (ɛ), area goodness factor (G) and efficiency index (η) and the lowest values of fluid-flow parameters like ( f) and hence the best efficiency, are achieved for -15° CFUp down-stream, ("case 3" of -15° CFUp down-stream and 6 mm LF height) and +5° CFDn down-stream. Correlations of Nu number, ( f) and (ɛ) as a function of θ and Re for the studied cases are performed.

  16. 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.

  17. SBWR core thermal hydraulic analysis during startup

    International Nuclear Information System (INIS)

    Lin, J.H.; Huang, R.L.; Sawyer, C.D.

    1993-01-01

    This paper reports on a thermal hydraulic analysis of the SIMPLIFIED BOILING WATER REACTOR (SBWR) during startup. The potential instability during a SBWR startup has drawn the attention of designers, researchers, and engineers. It has not been a concern for a Boiling Water Reactor (BWR) with forced recirculation; however, for SBWR with natural circulation the concern exists. The concern is about the possibility of a geysering mode oscillation during SBWR startup from a cold temperature and a low system pressure with a low natural circulation flow rate. A thermal hydraulic analysis of the SBWR is performed in simulation of the startup using the TRACG computer code. The temperature, pressure, and reactor power profiles of SBWR during the startup are presented. The results are compared with the data of a natural circulation boiling water reactor, the DODEWAARD plant, in which no instabilities have been observed during many startups. It is shown that a SBWR startup which follows proper procedures, geysering and other modes of oscillations can be avoided

  18. 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.

  19. Assessment of the impact of neutronic/thermal-hydraulic coupling on the design and performance of nuclear reactors for space propulsion

    International Nuclear Information System (INIS)

    Aithal, S.M.; Aldemir, T.; Vafai, K.

    1994-01-01

    A series of studies has been performed to investigate the potential impact of the coupling between neutronics and thermal hydraulics on the design and performance assessment of solid core reactors for nuclear thermal space propulsion, using the particle bed reactor (PBR) concept as an example system. For a given temperature distribution in the reactor, the k eff and steady-state core power distribution are obtained from three-dimensional, continuous energy Monte Carlo simulations using the MCNP code. For a given core power distribution, determination of the temperature distribution in the core and hydrogen-filled annulus between the reflector and pressure vessel is based on a nonthermal equilibrium analysis. The results show that a realistic estimation of fuel, core size, and control requirements for PBRs using hydrogenous moderators, as well as optimization of the overall engine design, may require coupled neutronic/thermal-hydraulic studies. However, it may be possible to estimate the thermal safety margins and propellant exit temperatures based on power distributions obtained from neutronic calculations at room temperature. The results also show that, while variation of the hydrogen flow rate in the annulus has been proposed as a partial control mechanism for PBRs, such control mechanism may not be feasible for PBRs with high moderator-to-fuel ratios and hence soft core neutron spectra

  20. Comparative of the Tribological Performance of Hydraulic Cylinders Coated by the Process of Thermal Spray HVOF and Hard Chrome Plating

    Directory of Open Access Journals (Sweden)

    R.M. Castro

    2014-03-01

    Full Text Available Due to the necessity of obtaining a surface that is resistant to wear and oxidation, hydraulic cylinders are typically coated with hard chrome through the process of electroplating process. However, this type of coating shows an increase of the area to support sealing elements, which interferes directly in the lubrication of the rod, causing damage to the seal components and bringing oil leakage. Another disadvantage in using the electroplated hard chromium process is the presence of high level hexavalent chromium Cr+6 which is not only carcinogenic, but also extremely contaminating to the environment. Currently, the alternative process of high-speed thermal spraying (HVOF - High Velocity Oxy-Fuel, uses composite materials (metal-ceramic possessing low wear rates. Research has shown that some mechanical properties are changed positively with the thermal spray process in industrial applications. It is evident that a coating based on WC has upper characteristics as: wear resistance, low friction coefficient, with respect to hard chrome coatings. These characteristics were analyzed by optical microscopy, roughness measurements and wear test.

  1. Horizontal steam generator thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Ubra, O. [SKODA Praha Company, Prague (Czechoslovakia); Doubek, M. [Czech Technical Univ., Prague (Czechoslovakia)

    1995-09-01

    Horizontal steam generators are typical components of nuclear power plants with pressure water reactor type VVER. Thermal-hydraulic behavior of horizontal steam generators is very different from the vertical U-tube steam generator, which has been extensively studied for several years. To contribute to the understanding of the horizontal steam generator thermal-hydraulics a computer program for 3-D steady state analysis of the PGV-1000 steam generator has been developed. By means of this computer program, a detailed thermal-hydraulic and thermodynamic study of the horizontal steam generator PGV-1000 has been carried out and a set of important steam generator characteristics has been obtained. The 3-D distribution of the void fraction and 3-D level profile as functions of load and secondary side pressure have been investigated and secondary side volumes and masses as functions of load and pressure have been evaluated. Some of the interesting results of calculations are presented in the paper.

  2. Thermal hydraulic reactor safety analyses and experiments

    International Nuclear Information System (INIS)

    Holmstroem, H.; Eerikaeinen, L.; Kervinen, T.; Kilpi, K.; Mattila, L.; Miettinen, J.; Yrjoelae, V.

    1989-04-01

    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.)

  3. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 3: Thermal hydraulic research and codes; Digital instrumentation and control; Structural performance

    International Nuclear Information System (INIS)

    Monteleone, S.

    1998-04-01

    This three-volume report contains papers presented at the conference. 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, Japan, Norway, and Russia. 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. This volume contains the following: (1) thermal hydraulic research and codes; (2) digital instrumentation and control; (3) structural performance

  4. 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.

  5. Local study of flow and low Reynolds thermal-hydraulic performance of a corrugated plane duct: application to plate heat exchangers

    International Nuclear Information System (INIS)

    Hugonnot, Patrick

    1989-01-01

    This research thesis addresses the local study of a flow in a corrugated plane duct by using experimental and numerical approaches on the one hand, and the experimental determination of thermal-hydraulic performance at low Reynolds number of different plate heat exchanger ducts on the other hand. Experimental visualisations of the local flow allowed regime transitions in 2D and 3D geometries to be determined. As far as the 2D duct is concerned, a wave profile optimisation is proposed, and the numerical study performed by using the TRIO software is in good agreement with experimental results. The optimised duct configuration can thus be envisaged for an industrial development. The determination of the friction coefficient and of the global heat exchange coefficient of different corrugated ducts allows plate exchangers to be sized on a wide range of Reynolds numbers. The respective influences of natural convection and of fluid thermal dependency on heat exchange have been studied [fr

  6. Proceedings of the third nuclear thermal hydraulics meeting

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    This book contains the proceedings of the Thermal Hydraulics Division of the American Nuclear Society. The papers presented include: Simulator qualification using engineering codes and Development of thermal hydraulic analysis capabilities for Oyster Creek

  7. 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.

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

    International Nuclear Information System (INIS)

    Ubra, O.; Doubek, M.

    1995-01-01

    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.)

  9. 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)

    1996-12-31

    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.

  10. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Baek, Won Pil; Song, C. H.; Kim, Y. S.

    2007-02-01

    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

  11. 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.

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

    International Nuclear Information System (INIS)

    Bang, Young Seok; Lee, S. H.; Ryu, Y. H.

    2001-02-01

    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

  13. Analysis of uncertainties of thermal hydraulic calculations

    International Nuclear Information System (INIS)

    Macek, J.; Vavrin, J.

    2002-12-01

    In 1993-1997 it was proposed, within OECD projects, that a common program should be set up for uncertainty analysis by a probabilistic method based on a non-parametric statistical approach for system computer codes such as RELAP, ATHLET and CATHARE and that a method should be developed for statistical analysis of experimental databases for the preparation of the input deck and statistical analysis of the output calculation results. Software for such statistical analyses would then have to be processed as individual tools independent of the computer codes used for the thermal hydraulic analysis and programs for uncertainty analysis. In this context, a method for estimation of a thermal hydraulic calculation is outlined and selected methods of statistical analysis of uncertainties are described, including methods for prediction accuracy assessment based on the discrete Fourier transformation principle. (author)

  14. Thermal hydraulics in undergraduate nuclear engineering education

    International Nuclear Information System (INIS)

    Theofanous, T.G.

    1986-01-01

    The intense safety-related research efforts of the seventies in reactor thermal hydraulics have brought about the recognition of the subject as one of the cornerstones of nuclear engineering. Many nuclear engineering departments responded by building up research programs in this area, and mostly as a consequence, educational programs, too. Whether thermal hydraulics has fully permeated the conscience of nuclear engineering, however, remains yet to be seen. The lean years that lie immediately ahead will provide the test. The purpose of this presentation is to discuss the author's own educational activity in undergraduate nuclear engineering education over the past 10 yr or so. All this activity took place at Purdue's School of Nuclear Engineering. He was well satisfied with the results and expects to implement something similar at the University of California in Santa Barbara in the near future

  15. 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 ...

  16. A Study on thermal-hydraulic characteristics of the coolant materials for the transmutation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hyun; You, Young Woo; Cho, Jae Seon; Kim, Ju Youl; Kim, Do Hyoung; Kim, Yoon Ik; Yang, Hui Chang [Seoul National University, Taejon (Korea)

    1998-03-01

    The objective of this study is to provide the direction of transmutation reactor design in terms of thermal hydraulics especially through the analysis of thermal hydraulic characteristics of various candidate materials for the transmutation reactor coolant. In this study, the characteristics of coolant materials used in current nuclear power plants and candidate materials for transmutation reactor are analyzed and compared. To evaluate the thermal hydraulic characteristics, the preliminary thermal-hydraulic calculation is performed for the candidate coolant materials of transmutation reactor. An analysis of thermal-hydraulic characteristics of transmutation reactor. An analysis of thermal-hydraulic characteristics of Sodium, Lead, Lead-Bismuth, and Lead-Lithium among the liquid metals considered as the coolant of transmutation reactor is performed by using computational fluid dynamics code FLUENT, and SIMPLER algorithm. (author). 50 refs., 40 figs., 30 tabs.

  17. Virginia Power thermal-hydraulics methods

    International Nuclear Information System (INIS)

    Anderson, R.C.; Basehore, K.L.; Harrell, J.R.

    1987-01-01

    Virginia Power's nuclear safety analysis group is responsible for the safety analysis of reload cores for the Surry and North Anna power stations, including the area of core thermal-hydraulics. Postulated accidents are evaluated for potential departure from nucleate boiling violations. In support of these tasks, Virginia Power has employed the COBRA code and the W-3 and WRB-1 DNB correlations. A statistical DNBR methodology has also been developed. The code, correlations and statistical methodology are discussed

  18. Thermal-Hydraulic Experiment Facility (THEF)

    International Nuclear Information System (INIS)

    Martinell, J.S.

    1982-01-01

    This paper provides an overview of the Thermal-Hydraulic Experiment Facility (THEF) at the Idaho National Engineering Laboratory (INEL). The overview describes the major test systems, measurements, and data acquisition system, and presents objectives, facility configuration, and results for major experimental projects recently conducted at the THEF. Plans for future projects are also discussed. The THEF is located in the Water Reactor Research Test Facility (WRRTF) area at the INEL

  19. Thermal-hydraulics of actinide burner reactors

    International Nuclear Information System (INIS)

    Takizuka, Takakazu; Mukaiyama, Takehiko; Takano, Hideki; Ogawa, Toru; Osakabe, Masahiro.

    1989-07-01

    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)

  20. Thermal-Hydraulic Tests for Reactor Core Safety

    International Nuclear Information System (INIS)

    Chun, Se Young; Chung, Moon Ki; Baek, Won Pil and others

    2005-04-01

    The reflood experiments for single rod annulus geometry have been performed to investigate the effect of spacer grid on thermal-hydraulics under reflood conditions. The reflood experimental loop for 6x6 rod bundle with a spacer grid developed in Korea has been provided. About 8000 data points for Post-CHF heat transfer have been obtained from the experiments About 1400 CHF data points for 3x3 Water and 5x5 Freon rod bundles have been obtained. The existing evaluation methodology for core safety under return-to-power conditions has been investigated using KAERI low flow CHF database. The hydraulic tests for turbulence mixing characteristics in subchannel of 5x5 rod bundle have been carried out using advanced measurement technique, LVD and the database for various spacer grids have been provided. In order to measure the turbulence mixing characteristics in details, the hydraulic loop with a magnified 5x5 rod bundle has been prepared. The database which was constructed through a systematic thermal hydraulic tests for the reflood phenomenon, CHF, Post-CHF is surely to be useful to the industry field, the regulation body and the development of thermal-hydraulic analysis code

  1. Thermal-Hydraulics analysis of pressurized water reactor core by using single heated channel model

    Directory of Open Access Journals (Sweden)

    Reza Akbari

    2017-08-01

    Full Text Available Thermal hydraulics of nuclear reactor as a basis of reactor safety has a very important role in reactor design and control. The thermal-hydraulic analysis provides input data to the reactor-physics analysis, whereas the latter gives information about the distribution of heat sources, which is needed to perform the thermal-hydraulic analysis. In this study single heated channel model as a very fast model for predicting thermal hydraulics behavior of pressurized water reactor core has been developed. For verifying the results of this model, we used RELAP5 code as US nuclear regulatory approved thermal hydraulics code. The results of developed single heated channel model have been checked with RELAP5 results for WWER-1000. This comparison shows the capability of single heated channel model for predicting thermal hydraulics behavior of reactor core.

  2. Thermal-hydraulic codes validation for safety analysis of NPPs with RBMK

    International Nuclear Information System (INIS)

    Brus, N.A.; Ioussoupov, O.E.

    2000-01-01

    This work is devoted to validation of western thermal-hydraulic codes (RELAP5/MOD3 .2 and ATHLET 1.1 Cycle C) in application to Russian designed light water reactors. Such validation is needed due to features of RBMK reactor design and thermal-hydraulics in comparison with PWR and BWR reactors, for which these codes were developed and validated. These validation studies are concluded with a comparison of calculation results of modeling with the thermal-hydraulics codes with the experiments performed earlier using the thermal-hydraulics test facilities with the experimental data. (authors)

  3. Development of thermal hydraulic evaluation code for CANDU reactors

    International Nuclear Information System (INIS)

    Kim, Man Woong; Yu, Seon Oh; Choi, Yong Seog; Shin, Chull; Hwang, Soo Hyun

    2004-02-01

    To enhance the safety of operating CANDU reactors, the establishment of the safety analysis codes system for CANDU reactors is in progress. As for the development of thermal-hydraulic analysis code for CANDU system, the studies for improvement of evaluation model inside RELAP/CANDU code and the development of safety assessment methodology for GSI (Generic Safety Issues) are in progress as a part of establishment of CANDU safety assessment system. To develop the 3-D thermal-hydraulic analysis code for moderator system, the CFD models for analyzing the CANDU-6 moderator circulation are developed. One model uses a structured grid system with the porous media approach for the 380 Calandria tubes in the core region. The other uses a unstructured grid system on the real geometry of 380 Calandria tubes, so that the detailed fluid flow between the Calandria tubes can be observed. As to the development of thermal-hydraulic analysis code for containment, the study on the applicability of CONTAIN 2.0 code to a CANDU containment was conducted and a simulation of the thermal-hydraulic phenomena during the accident was performed. Besides, the model comparison of ESFs (Engineered Safety Features) inside CONTAIN 2.0 code and PRESCON code has also conducted

  4. Development of design technology on thermal-hydraulic performance in tight-lattice rod bundles. II-rod bowing effect on boiling transition

    International Nuclear Information System (INIS)

    Liu, Wei; Tamai, Hidesada; Kureta, Masatoshi; Ohnuki, Akira; Takase, Kazuyuki; Akimoto, Hajime

    2007-01-01

    A thermal-hydraulic feasibility project for an Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed since 2002. In this R and D project, large-scale thermal-hydraulic tests, several model experiments and development of advanced numerical analysis codes have been carried out. In this paper, we will describe the critical power characteristics in a 37-rod tight-lattice bundle with rod-bowing under both steady and transient states. It is observed that no matter it is run under a steady or a transient state, boiling transition (BT) always occurs axially at exit elevation of upper high-heat-flux region and transversely in the central area of the bundle. Steady critical power increases monotonically with the increase of mass velocity, with the decrease of inlet water temperature and with the decrease of exit pressure. These trends are same as those in the base case test without rod-bowing. The steady critical power with rod-bowing is about 10% lower than that without rod-bowing. For the postulated power increase and flow decrease cases that may be possibly met in a normal operation of the FLWR, it is confirmed that no BT occurs when Initial Critical Power Ratio (ICPR) is 1.3. Moreover, when the transitions are run under severer ICPR that causes BT, the transient critical powers are generally same as the steady ones. The experiments are analyzed with TRAC-BF1 code. The TRAC-BF1 code shows good prediction for the occurrence or the non occurrence of the BT and predicts the BT starting time within the accuracy of critical power correlation. Traditional quasi - steady state prediction of the transient BT is confirmed being applicable for the postulated abnormal transient processes in the tight lattice bundle with rod - bowing. (author)

  5. Review of computational thermal-hydraulic modeling

    International Nuclear Information System (INIS)

    Keefer, R.H.; Keeton, L.W.

    1995-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-08-15

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

  7. Mercury Thermal Hydraulic Loop (MTHL) Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Felde, David K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Crye, Jason Michael [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wendel, Mark W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yoder, Jr, Graydon L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Farquharson, George [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jallouk, Philip A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McFee, Marshall T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pointer, William David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ruggles, Art E. [Univ. of Tennessee, Knoxville, TN (United States); Carbajo, Juan J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-03-01

    The Spallation Neutron Source (SNS) is a high-power linear accelerator built at Oak Ridge National Laboratory (ORNL) which incorporates the use of a flowing liquid mercury target. The Mercury Thermal Hydraulic Loop (MTHL) was constructed to investigate and verify the heat transfer characteristics of liquid mercury in a rectangular channel. This report provides a compilation of previously reported results from the water-cooled and electrically heated straight and curved test sections that simulate the geometry of the window cooling channel in the target nose region.

  8. Applied mathematical methods in nuclear thermal hydraulics

    International Nuclear Information System (INIS)

    Ransom, V.H.; Trapp, J.A.

    1983-01-01

    Applied mathematical methods are used extensively in modeling of nuclear reactor thermal-hydraulic behavior. This application has required significant extension to the state-of-the-art. The problems encountered in modeling of two-phase fluid transients and the development of associated numerical solution methods are reviewed and quantified using results from a numerical study of an analogous linear system of differential equations. In particular, some possible approaches for formulating a well-posed numerical problem for an ill-posed differential model are investigated and discussed. The need for closer attention to numerical fidelity is indicated

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

    International Nuclear Information System (INIS)

    Heard, F.J.; Cramer, E.R.; Beaver, T.R.; Thurgood, M.J.

    1996-01-01

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

    International Nuclear Information System (INIS)

    Ishikawa, M.; Kitamura, H.; Toba, A.; Omoto, A.

    2004-01-01

    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)

  11. Thermal-hydraulics of helium cooled First Wall channels and scoping investigations on performance improvement by application of ribs and mixing devices

    Energy Technology Data Exchange (ETDEWEB)

    Arbeiter, Frederik, E-mail: frederik.arbeiter@kit.edu [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, Christian [EUROfusion – Programme Management Unit, Garching (Germany); Chen, Yuming; Ilić, Milica; Schwab, Florian [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Sieglin, Bernhard [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Wenninger, Ronald [EUROfusion – Programme Management Unit, Garching (Germany)

    2016-11-01

    Highlights: • Existing first wall designs and expected plasma heat loads are reviewed. • Heat transfer enhancement methods are investigated by CFD. • The results for heat transfer and friction are given, compared and explained. • Relations for needed pumping power and gained thermal heat are shown. • A range for the maximum permissible heat loads from the plasma is estimated. - Abstract: The first wall (FW) of DEMO is a component with high thermal loads. The cooling of the FW has to comply with the material's upper and lower temperature limits and requirements from stress assessment, like low temperature gradients. Also, the cooling has to be integrated into the balance-of-plant, in a sense to deliver exergy to the power cycle and require a limited pumping power for coolant circulation. This paper deals with the basics of FW cooling and proposes optimization approaches. The effectiveness of several heat transfer enhancement techniques is investigated for the use in helium cooled FW designs for DEMO. Among these are wall-mounted ribs, large scale mixing devices and modified hydraulic diameter. Their performance is assessed by computational fluid dynamics (CFD), and heat transfer coefficients and pressure drop are compared. Based on the results, an extrapolation to high heat fluxes is tried to estimate the higher limits of cooling capabilities.

  12. Thermal-hydraulic analysis of heat transfer in subchannels of the European high performance supercritical Water-Cooled Reactor for different CFD turbulence models

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Landy Y.; Rojas, Leorlen Y.; Gamez, Abel; Rosales, Jesus; Gonzalez, Daniel; Garcia, Carlos, E-mail: lcastro@instec.cu, E-mail: leored1984@gmail.com, E-mail: agamezgmf@gmail.com, E-mail: jrosales@instec.cu, E-mail: danielgonro@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Oliveira, Carlos Brayner de, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Dominguez, Dany S., E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Pos-Graduacao em Modelagem Computacional

    2015-07-01

    Chosen as one of six Generation‒IV nuclear-reactor concepts, Supercritical Water-cooled Reactors (SCWRs) are expected to have high thermal efficiencies within the range of 45 - 50% owing to the reactor's high pressures and outlet temperatures. In this reactor, the primary water enters the core under supercritical-pressure condition (25 MPa) at a temperature of 280 deg C and leaves it at a temperature of up to 510 deg C. Due to the significant changes in the physical properties of water at supercritical-pressure, the system is susceptible to local temperature, density and power oscillations. The behavior of supercritical water into the core of the SCWR, need to be sufficiently studied. Most of the methods available to predict the effects of the heat transfer phenomena within the pseudocritical region are based on empirical one-directional correlations, which do not capture the multidimensional effects and do not provide accurate results in regions such as the deteriorated heat transfer regime. In this paper, computational fluid dynamics (CFD) analysis was carried out to study the thermal-hydraulic behavior of supercritical water flows in sub-channels of a typical European High Performance Light Water Reactor (HPLWR) fuel assembly using commercial CFD code CFX-14. It was determined the steady-state equilibrium parameters and calculated the temperature and density distributions. A comparative study for different turbulence models were carried out and the obtained results are discussed. (author)

  13. Thermal-hydraulic analysis of heat transfer in subchannels of the European high performance supercritical Water-Cooled Reactor for different CFD turbulence models

    International Nuclear Information System (INIS)

    Castro, Landy Y.; Rojas, Leorlen Y.; Gamez, Abel; Rosales, Jesus; Gonzalez, Daniel; Garcia, Carlos; Oliveira, Carlos Brayner de; Dominguez, Dany S.

    2015-01-01

    Chosen as one of six Generation‒IV nuclear-reactor concepts, Supercritical Water-cooled Reactors (SCWRs) are expected to have high thermal efficiencies within the range of 45 - 50% owing to the reactor's high pressures and outlet temperatures. In this reactor, the primary water enters the core under supercritical-pressure condition (25 MPa) at a temperature of 280 deg C and leaves it at a temperature of up to 510 deg C. Due to the significant changes in the physical properties of water at supercritical-pressure, the system is susceptible to local temperature, density and power oscillations. The behavior of supercritical water into the core of the SCWR, need to be sufficiently studied. Most of the methods available to predict the effects of the heat transfer phenomena within the pseudocritical region are based on empirical one-directional correlations, which do not capture the multidimensional effects and do not provide accurate results in regions such as the deteriorated heat transfer regime. In this paper, computational fluid dynamics (CFD) analysis was carried out to study the thermal-hydraulic behavior of supercritical water flows in sub-channels of a typical European High Performance Light Water Reactor (HPLWR) fuel assembly using commercial CFD code CFX-14. It was determined the steady-state equilibrium parameters and calculated the temperature and density distributions. A comparative study for different turbulence models were carried out and the obtained results are discussed. (author)

  14. Thermal-hydraulic performance of a multiple jet cooling module with a concave dimple array in a helium-cooled divertor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hyo-Yeon; Kim, Kwang-Yong, E-mail: kykim@inha.ac.kr

    2017-01-15

    A numerical study was performed to evaluate the thermal-hydraulic performance of a finger type cooling module, where multiple jets impinge on the surface with concave dimples, in the divertor of a nuclear fusion reactor. Conjugate heat transfer was analyzed in both the solid and fluid domains using three-dimensional Reynolds-averaged Navier-Stokes equations with the shear stress transport turbulence model. The computational domain consisted of a single fluid domain and three solid domains: tile, thimble, and cartridge. The numerical results for the temperature variation on the tile were validated in comparison with the experimental data. A parametric study was performed with two design variables, the ratios of dimple diameter and dimple height to the nozzle diameter, and two dimple arrays (inline and staggered arrays). The parametric study showed that the heat transfer rate was increased by up to 2.62% by introducing concave dimples, and that the heat transfer and pressure drop performances increased with increasing diameter and height of the dimples for a specified dimple array.

  15. Thermal hydraulic design of PFBR core

    International Nuclear Information System (INIS)

    Roychowdhury, D.G.; Vinayagam, P.P.; Ravichandar, S.C.

    2000-01-01

    The thermal-hydraulic design of core is important in respecting temperature limits while achieving higher outlet temperature. This paper deals with the analytical process developed and implemented for analysing steady state thermal-hydraulics of PFBR core. A computer code FLONE has been developed for optimisation of flow allocation through the subassemblies (SA). By calibrating β n (ratio between the maximum channel temperature rise and SA average temperature rise) values with SUPERENERGY code and using these values in FLONE code, prediction of average and maximum coolant temperature distribution is found to be reasonably accurate. Hence, FLONE code is very powerful design tool for core design. A computer code SAPD has been developed to calculate the pressure drop of fuel and blanket SA. Selection of spacer wire pitch depends on the pressure drop, flow-induced vibration and the mixing characteristics. A parametric study was made for optimisation of spacer wire pitch for the fuel SA. Experimental programme with 19 pin-bundle has been undertaken to find the flow-induced vibration characteristics of fuel SA. Also, experimental programme has been undertaken on a full-scale model to find the pressure drop characteristics in unorificed SA, orifices and the lifting force on the SA. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M H; Hur, C; Kim, D K; Cho, H J [POhang Univ., of Science and TECHnology, Pohang (Korea, Republic of)

    1996-07-15

    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.

  17. Experimental Investigation on Effect of Fin Shape on the Thermal-Hydraulic Performance of Compact Fin-and-Tube Heat Exchangers

    Science.gov (United States)

    Moorthy, P.; Oumer, A. N.; Ishak, M.

    2018-03-01

    The aim of this paper is to investigate the effect of fin shapes on the performance of compact finned flat tube heat exchangers. Three types of fin shapes namely plain, wavy, and rectangular grooved fins attached to three by three arrays of flat tube banks were considered. Moreover, the tubes were deployed in in-line and staggered arrangements. In addition to the fin shapes, the air velocity and the tube inclination angles were varied and the thermal-hydraulic performance was analysed. On the other hand, the temperatures at the tube surfaces were kept constant to produce constant heat flux throughout the study. The results showed that as flowrate increases, the heat transfer increases, however, the friction factor decreases. Staggered arrangement produces higher heat transfer and friction factor than inline fin. Moreover, the rectangular fin is the best in terms of high heat transfer however the drawback of high friction factor leads the fin to have the least efficiency of all. On the other hand, plain fin had the least heat transfer performance however the highest efficiency was achieved. Therefore, plain fin should be used when efficiency is prioritized and rectangular fin when high heat transfer is desired.

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

    International Nuclear Information System (INIS)

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

    2005-04-01

    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

  19. 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.

  20. BWR 9 X 9 Fuel Assembly Thermal-Hydraulic Tests (2): Hydraulic Vibration Test

    International Nuclear Information System (INIS)

    Yoshiaki Tsukuda; Katsuichiro Kamimura; Toshiitsu Hattori; Akira Tanabe; Noboru Saito; Masahiko Warashina; Yuji Nishino

    2002-01-01

    Nuclear Power Engineering Corporation (NUPEC) conducted thermal-hydraulic projects for verification of thermal-hydraulic design reliability for BWR high-burnup 8 x 8 and 9 x 9 fuel assemblies, entrusted by the Ministry of Economy, Trade and Industry (METI). As a part of the NUPEC thermal-hydraulic projects, hydraulic vibration tests using full-scale test assemblies simulating 9 x 9 fuel assemblies were carried out to evaluate BWR fuel integrity. The test data were applied to development of a new correlation for the estimation of fuel rod vibration amplitude. (authors)

  1. VHTR core modeling: coupling between neutronic and thermal-hydraulics

    International Nuclear Information System (INIS)

    Limaiem, I.; Damian, F.; Raepsaet, X.; Studer, E.

    2005-01-01

    Following the present interest in the next generation nuclear power plan (NGNP), Cea is deploying special effort to develop new models and qualify its research tools for this next generation reactors core. In this framework, the Very High Temperature Reactor concept (VHTR) has an increasing place in the actual research program. In such type of core, a strong interaction exists between neutronic and thermal-hydraulics. Consequently, the global core modelling requires accounting for the temperature feedback in the neutronic models. The purpose of this paper is to present the new neutronic and thermal-hydraulics coupling model dedicated to the High Temperature Reactors (HTR). The coupling model integrates a new version of the neutronic scheme calculation developed in collaboration between Cea and Framatome-ANP. The neutronic calculations are performed using a specific calculation processes based on the APOLLO2 transport code and CRONOS2 diffusion code which are part of the French reactor physics code system SAPHYR. The thermal-hydraulics model is characterised by an equivalent porous media and 1-D fluid/3-D thermal model implemented in the CAST3M/ARCTURUS code. The porous media approach involves the definition of both homogenous and heterogeneous models to ensure a correct temperature feedback. This study highlights the sensitivity of the coupling system's parameters (radial/axial meshing and data exchange strategy between neutronic and thermal-hydraulics code). The parameters sensitivity study leads to the definition of an optimal coupling system specification for the VHTR. Besides, this work presents the first physical analysis of the VHTR core in steady-state condition. The analysis gives information about the 3-D power peaking and the temperature coefficient. Indeed, it covers different core configurations with different helium distribution in the core bypass. (authors)

  2. Thermal hydraulic simulation of the CANDU nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, Athos M.S.S. de; Ramos, Mario C.; Costa, Antonella L.; Fernandes, Gustavo H.N., E-mail: athos1495@yahoo.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Rio de janeiro, RJ (Brazil)

    2017-07-01

    The CANDU (Canada Deuterium Uranium) is a Canadian-designed power reactor of PHWR type (Pressurized Heavy Water Reactor) that uses heavy water (deuterium oxide) for moderator and coolant, and natural uranium for fuel. There are about 47 reactors of this type in operation around the world generating more than 23 GWe, highlighting the importance of this kind of device. In this way, the main purpose of this study is to develop a thermal hydraulic model for a CANDU reactor to aggregate knowledge in this line of research. In this way, a core modeling was performed using RELAP5-3D code. Results were compared with reference data to verify the model behavior in steady state operation. Thermal hydraulic parameters as temperature, pressure and mass flow rate were verified and the results are in good agreement with reference data, as it is being presented in this work. (author)

  3. Thermal-hydraulic tests for reactor safety system

    International Nuclear Information System (INIS)

    Chun, Se Young; Chung, Moon Ki; Baek, Won Pil

    2002-05-01

    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

  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. 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...

  6. An overview on rod-bundle thermal-hydraulic analyses

    International Nuclear Information System (INIS)

    Sha, W.T.

    1980-01-01

    Three methods used in rod-bundle thermal-hydraulic analysis are summarized. These methods are: (1) subchannel analysis, (2) porous medium formulation with volume porosity, surface permeability, distributed resistance and distributed heat source (sink) and, (3) bench-mark rod-bundle thermal-hydraulic analysis using a boundary-fitted coordinate system. Basic limitations and merits of each method are delineated. (orig.)

  7. CFD studies on thermal hydraulics of spallation targets

    International Nuclear Information System (INIS)

    Tak, N.I.; Batta, A.; Cheng, X.

    2005-01-01

    Full text of publication follows: Due to the fast advances in computer hardware as well as software in recent years, more and more interests have been aroused to use computational fluid dynamics (CFD) technology in nuclear engineering and designs. During recent many years, Forschungszentrum Karlsruhe (FZK) has been actively involved in the thermal hydraulic analysis and design of spallation targets. To understand the thermal hydraulic behaviors of spallation targets very detailed simulations are necessary because of their complex geometries, complicated boundary conditions such as spallation heat distributions, and very strict design limits. A CFD simulation is believed to be the best for this purpose even though the validation of CFD codes are not perfectly completed yet in specific topics like liquid metal heat transfer. The research activities on three spallation targets (i.e., MEGAPIE, TRADE, and XADS targets) are currently very active in Europe in order to consolidate the European ADS road-map. In the thermal hydraulics point of view, two kinds of the research activities, i.e., (1) numerical design and (2) experimental work, are required to achieve the objectives of these targets. It should be noted that CFD studies play important role on both kinds of two activities. A preliminary design of a target can be achieved by sophisticated CFD analysis and pre-and-post analyses of an experimental work using a CFD code help the design of the test section of the experiment as well as the analysis of the experimental results. The present paper gives an overview about the recent CFD studies relating to thermal hydraulics of the spallation targets recently involved in FZK. It covers numerical design studies as well as CFD studies to support experimental works. The CFX code has been adopted for the studies. Main recent results for the selected examples performed by FZK are presented and discussed with their specific lessons learned. (authors)

  8. Evaluation of thermal-hydraulic performance of hydrocarbon refrigerants during flow boiling in a microchannels array heat sink

    International Nuclear Information System (INIS)

    Chávez, Cristian A.; Leão, Hugo L.S.L.; Ribatski, Gherhardt

    2017-01-01

    Highlights: • Evaluation of refrigerants R600a, R290 and R1270 during flow boiling in a microchannels array. • Comparison of data for hydrocarbons with previous data for R134a. • Parametric analysis of heat transfer coefficient, pressure drop, ONB and exergy behaviors. • Comparison of the experimental data and prediction methods from literature. • In general, refrigerant R290 presents the best performance. - Abstract: The present study concerns an experimental evaluation of the performance of hydrocarbon refrigerants during flow boiling in a microchannels array heat sink. The heat sink is composed of fifty channels with cross sectional areas of 123 × 494 μm"2 and length of 15 mm manufactured in a copper block. Heat transfer coefficient and pressure drop data were obtained for refrigerants R600a, R290 and R1270, mass velocities from 165 to 823 kg/m"2 s, heat fluxes up to 400 kW/m"2, liquid subcooling at the inlet of the test section of 5, 10 and 15 °C and saturation temperature of 25 °C. The data were compared with experimental results obtained in a previous study for R134a and predictions by methods from literature. In general, R290 presented the best performance, providing the highest average heat transfer coefficient and a pressure drop only slightly higher than R1270 that was the fluid presenting the lowest pressure drop. An exergy analysis also revealed the refrigerant R290 as the one presenting the best performance. However, R290 needed the highest excess of superheating to trigger the boiling process (ONB). The methods from literature evaluated in the present study poorly predicted the experimental data for two-phase pressure drop. On the other hand, the method of Kanizawa et al. (2016) was quite accurate in predicting the heat transfer results.

  9. Experimental investigation of thermal-hydraulic performance of PCCS with horizontal tube heat exchangers: single U-tube test

    International Nuclear Information System (INIS)

    Nakamura, Hideo; Anoda, Yoshinari; Arai, Kenji; Kurita, Tomohisa

    2000-01-01

    JAERI and JAPC started a cooperative study to verify performance of a PCCS (Passive Containment Cooling System) using horizontal heat exchanger for next-generation BWR in 1998. A test facility with a horizontal single U-tube was constructed in JAERI in 1999 to investigate fundamental condensation behavior under influences of non-condensable gas. Preliminary pre-test analyses were performed using RELAP5/ MOD3.2.1.2 code to expect the experimental outcomes by incorporating a correlation for condensation degradation because of non-condensable gas by Ueno et al. for better prediction. Preliminary results from both experiments (shakedown) and pre-test analyses indicated that the PCCS using horizontal U-tube heat exchanger is promising. Steam generated under assumed severe accident conditions; steam generation rate approx. = 1% core power, non-condensable gas concentration of 1% and simulated containment vessel pressure of 0.7 MPa, was totally condensed with a small differential pressure across inlet and outlet plenum. Experimental data will be accumulated to develop models and correlations for a better prediction of responses of the PCCS using horizontal heat exchanger during postulated severe accidents. (author)

  10. BEPU-FSAR: establishing a background for extension of nuclear thermal hydraulic principles to non thermal-hydraulic codes

    Energy Technology Data Exchange (ETDEWEB)

    Menzel, Francine; Sabundjian, Gaianê, E-mail: franmenzel@gmail.com, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); D’Auria, Francesco, E-mail: f.dauria@ing.unipi.it [University of Pisa, San Piero a Grado Nuclear Research Group (Italy)

    2017-07-01

    Nuclear thermal hydraulic and accident analysis are based in three pillar activities, which consists in: Scaling, Coupling and V and V. Each of them are established technology, with key documents to describe and widely used. The final goal of this work is to apply the BEPU methodology in all parts of FSAR where analytical techniques are needed (BEPU-FSAR) and for that the crucial step is the transfer of the BEPU concepts into the other areas. In this sense, the issue is how to adapt to other disciplines the pillar activities presented in the thermal hydraulic area. For that we need to identify which elements can be applied in the other areas, to show that the proposed methodology is feasible. This work aims to discuss the first steps towards a BEPU-FSAR methodology and to show that the Scaling, Coupling and V and V elements, currently done for thermal-hydraulic codes, can be also done for different codes, which are used to perform different analysis included on a FSAR of a generic plant. (author)

  11. 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)

  12. Investigating the thermal hydraulic performance of spacer grid with mixing vanes using STAR-CCM+ and MATRA

    International Nuclear Information System (INIS)

    Agbodemegbe, V. Y.

    2014-07-01

    Enhancement of heat transfer for flow through rod bundles is linked to the extent and sustainability of mixing in the flow geometry. Spacer grids used as support for rod bundles in nuclear reactors, when attached with ditferent designs of mixing vanes promote turbulent mixing by inducing swirl or forced lateral convection that improves mixing within or between sub-channels. The improved turbulent mixing raises the margin of the onset of critical heat flux in light water reactors (LWR) and also ensures a higher fuel cycle economy. To optimize design of mixing vanes and performance of spacer grids with mixing vanes, computational fluid dynamic simulations arc carried out on new designs and validated experimentally prior to industrial application. In the present study, computational fluid dynamic simulation using STAR-CCM+ and sub-channel analysis with MATRA were performed for flow of water through a 5 X 5 rod bundle geometry for which the rod to rod pitch to diameter ratio was 1.33 and the wall to rod pitch to diameter ratio was 0.74. The two layer k-epsilon turbulence model with an all- y + automatic wall treatment function in STAR-CCM+ were adopted for an isothermal single phase flow through the geometry with imposed cyclic periodic and non-cyclic periodic interface boundary conditions. The objective was to primarily investigate the detail flow behavior in rod bundle in the presence of spacer grids with and without attached mixing vanes. Furthermore, the present study also investigated the extent of turbulent mixing and lateral mass flux induced by the mixing vanes through comparative analysis of parametric trends for spacer grid with and without mixing vanes. New models and empirical correlations for describing the mixing vanes effects were also developed. These semi-empirical correlations improved the prediction for lateral mass flux due to turbulence, fraction of flow diverted through gaps and cross-flow resistance coefficients. Validation of simulation results

  13. Isotope Production Facility Conceptual Thermal-Hydraulic Design Review and Scoping Calculations

    International Nuclear Information System (INIS)

    Pasamehmetoglu, K.O.; Shelton, J.D.

    1998-01-01

    The thermal-hydraulic design of the target for the Isotope Production Facility (IPF) is reviewed. In support of the technical review, scoping calculations are performed. The results of the review and scoping calculations are presented in this report

  14. Current Development and Trends in Thermal-Hydraulics

    International Nuclear Information System (INIS)

    Toth, I.

    2008-01-01

    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.

  15. Thermal-hydraulic design of the 200 MW NHR

    International Nuclear Information System (INIS)

    Li Jincai; Gao Zuying; Xu Baocheng; He Junxiao

    1997-01-01

    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

  16. Thermal-hydraulic design of the 200 MW NHR

    Energy Technology Data Exchange (ETDEWEB)

    Jincai, Li; Zuying, Gao; Baocheng, Xu; Junxiao, He [Institute of Nuclear Energy and Technology, Tsingua Univ., Beijing (China)

    1997-09-01

    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-hydraulic characteristic of the PGV-1000 steam generator

    International Nuclear Information System (INIS)

    Ubra, O.; Doubek, M.

    1995-01-01

    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)

  18. Process management using component thermal-hydraulic function classes

    Science.gov (United States)

    Morman, J.A.; Wei, T.Y.C.; Reifman, J.

    1999-07-27

    A process management expert system where following malfunctioning of a component, such as a pump, for determining system realignment procedures such as for by-passing the malfunctioning component with on-line speeds to maintain operation of the process at full or partial capacity or to provide safe shut down of the system while isolating the malfunctioning component. The expert system uses thermal-hydraulic function classes at the component level for analyzing unanticipated as well as anticipated component malfunctions to provide recommended sequences of operator actions. Each component is classified according to its thermal-hydraulic function, and the generic and component-specific characteristics for that function. Using the diagnosis of the malfunctioning component and its thermal hydraulic class, the expert system analysis is carried out using generic thermal-hydraulic first principles. One aspect of the invention employs a qualitative physics-based forward search directed primarily downstream from the malfunctioning component in combination with a subsequent backward search directed primarily upstream from the serviced component. Generic classes of components are defined in the knowledge base according to the three thermal-hydraulic functions of mass, momentum and energy transfer and are used to determine possible realignment of component configurations in response to thermal-hydraulic function imbalance caused by the malfunctioning component. Each realignment to a new configuration produces the accompanying sequence of recommended operator actions. All possible new configurations are examined and a prioritized list of acceptable solutions is produced. 5 figs.

  19. Process management using component thermal-hydraulic function classes

    Science.gov (United States)

    Morman, James A.; Wei, Thomas Y. C.; Reifman, Jaques

    1999-01-01

    A process management expert system where following malfunctioning of a component, such as a pump, for determining system realignment procedures such as for by-passing the malfunctioning component with on-line speeds to maintain operation of the process at full or partial capacity or to provide safe shut down of the system while isolating the malfunctioning component. The expert system uses thermal-hydraulic function classes at the component level for analyzing unanticipated as well as anticipated component malfunctions to provide recommended sequences of operator actions. Each component is classified according to its thermal-hydraulic function, and the generic and component-specific characteristics for that function. Using the diagnosis of the malfunctioning component and its thermal hydraulic class, the expert system analysis is carried out using generic thermal-hydraulic first principles. One aspect of the invention employs a qualitative physics-based forward search directed primarily downstream from the malfunctioning component in combination with a subsequent backward search directed primarily upstream from the serviced component. Generic classes of components are defined in the knowledge base according to the three thermal-hydraulic functions of mass, momentum and energy transfer and are used to determine possible realignment of component configurations in response to thermal-hydraulic function imbalance caused by the malfunctioning component. Each realignment to a new configuration produces the accompanying sequence of recommended operator actions. All possible new configurations are examined and a prioritized list of acceptable solutions is produced.

  20. Hydraulic and thermal design of a gas microchannel heat exchanger

    International Nuclear Information System (INIS)

    Yang Yahui; Brandner, Juergen J; Morini, Gian Luca

    2012-01-01

    In this paper investigations on the design of a gas flow microchannel heat exchanger are described in terms of hydrodynamic and thermal aspects. The optimal choice for thermal conductivity of the solid material is discussed by analysis of its influences on the thermal performance of a micro heat exchanger. Two numerical models are built by means of a commercial CFD code (Fluent). The simulation results provide the distribution of mass flow rate, inlet pressure and pressure loss, outlet pressure and pressure loss, subjected to various feeding pressure values. Based on the thermal and hydrodynamic analysis, a micro heat exchanger made of polymer (PEEK) is designed and manufactured for flow and heat transfer measurements in air flows. Sensors are integrated into the micro heat exchanger in order to measure the local pressure and temperature in an accurate way. Finally, combined with numerical simulation, an operating range is suggested for the present micro heat exchanger in order to guarantee uniform flow distribution and best thermal and hydraulic performances.

  1. Thermal hydraulics analysis of LIBRA-SP target chamber

    International Nuclear Information System (INIS)

    Mogahed, E.A.

    1996-01-01

    LIBRA-SP is a conceptual design study of an inertially confined 1000 MWe fusion power reactor utilizing self-pinched light ion beams. There are 24 ion beams which are arranged around the reactor cavity. The reaction chamber is an upright cylinder with an inverted conical roof resembling a mushroom, and a pool floor. The vertical sides of the cylinder are occupied by a blanket zone consisting of many perforated rigid HT-9 ferritic steel tubes called PERITs (PEr-forated RIgid Tube). The breeding/cooling material, liquid lead-lithium, flows through the PERITs, providing protection to the reflector/vacuum chamber so as to make it a lifetime component. The neutronics analysis and cavity hydrodynamics calculations are performed to account for the neutron heating and also to determine the effects of vaporization/condensation processes on the surface heat flux. The steady state nuclear heating distribution at the midplane is used for thermal hydraulics calculations. The maximum surface temperature of the HT-9 is chosen to not exceed 625 degree C to avoid drastic deterioration of the metal's mechanical properties. This choice restricts the thermal hydraulics performance of the reaction cavity. The inlet first surface coolant bulk temperature is 370 degree C, and the heat exchanger inlet coolant bulk temperature is 502 degree C. 4 refs., 6 figs., 2 tabs

  2. Overview of the ANL advanced LMR system thermal-hydraulic test program supporting both GE/PRISM and RI/SAFR

    International Nuclear Information System (INIS)

    Oras, J.J.; Kuzay, T.M.; Kasza, K.E.

    1988-01-01

    Descriptions of the ANL thermal-hydraulic water models of both the PRISM and SAFR reactors are presented, together with results from Phases I and II of the thermal-hydraulic test program. Phenomena discovered during these tests and modeling results are presented. Overall, these efforts demonstrate the acceptable thermal-hydraulic performance of both the PRISM and SAFR concepts

  3. Progress of the DUPIC fuel compatibility analysis (II) - thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Choi, Hang Bok

    2005-03-01

    Thermal-hydraulic compatibility of the DUPIC fuel bundle with a 713 MWe Canada deuterium uranium (CANDU-6) reactor was studied by using both the single channel and sub-channel analysis methods. The single channel analysis provides the fuel channel flow rate, pressure drop, critical channel power, and the channel exit quality, which are assessed against the thermal-hydraulic design requirements of the CANDU-6 reactor. The single channel analysis by the NUCIRC code showed that the thermal-hydraulic performance of the DUPIC fuel is not different from that of the standard CANDU fuel. Regarding the local flow characteristics, the sub-channel analysis also showed that the uncertainty of the critical channel power calculation for the DUPIC fuel channel is very small. As a result, both the single and sub-channel analyses showed that the key thermal-hydraulic parameters of the DUPIC fuel channel do not deteriorate compared to the standard CANDU fuel channel.

  4. Thermal hydraulic evaluation of advanced wire-wrapped assemblies

    International Nuclear Information System (INIS)

    Wei, J.P.

    1975-01-01

    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

  5. Spent nuclear fuel storage pool thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Gay, R.R.

    1984-01-01

    Storage methods and requirements for spent nuclear fuel at U.S. commercial light water reactors are reviewed in Section 1. Methods of increasing current at-reactor storage capabilities are also outlined. In Section 2 the development of analytical methods for the thermal-hydraulic analysis of spent fuel pools is chronicled, leading up to a discussion of the GFLOW code which is described in Section 3. In Section 4 the verification of GFLOW by comparisons of the code's predictions to experimental data taken inside the fuel storage pool at the Maine Yankee nuclear power plant is presented. The predictions of GFLOW using 72, 224, and 1584 node models of the storage pool are compared to each other and to the experimental data. An example of thermal licensing analysis for Maine Yankee using the GFLOW code is given in Section 5. The GFLOW licensing analysis is compared to previous licensing analysis performed by Yankee Atomic using the RELAP-4 computer code

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    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.

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

    International Nuclear Information System (INIS)

    2004-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    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.

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

    International Nuclear Information System (INIS)

    2004-01-01

    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

  10. Thermal-hydraulic modeling needs for passive reactors

    International Nuclear Information System (INIS)

    Kelly, J.M.

    1997-01-01

    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

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

    International Nuclear Information System (INIS)

    Bokhari, I.H.; Israr, M.; Pervez, S.

    1999-01-01

    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 m 3 /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)

  12. 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.

  13. Finite volume thermal-hydraulics and neutronics coupled calculations - 15300

    International Nuclear Information System (INIS)

    Araujo Silva, V.; Campagnole dos Santos, A.A.; Mesquit, A.Z.; Bernal, A.; Miro, R.; Verdu, G.; Pereira, C.

    2015-01-01

    The computational power available nowadays allows the coupling of neutronics and thermal-hydraulics codes for reactor studies. The present methodology foresees at least one constraint to the separated codes in order to perform coupled calculations: both codes must use the same geometry, however, meshes can be different for each code as long as the internal surfaces stays the same. Using the finite volume technique, a 3D diffusion nodal code was implemented to deal with neutron transport. This code can handle non-structured meshes which allows for complicated geometries calculations and therefore more flexibility. A computational fluid dynamics (CFD) code was used in order to obtain the same level of details for the thermal hydraulics calculations. The chosen code is OpenFOAM, an open-source CFD tool. Changes in OpenFOAM allow simple coupled calculations of a PWR fuel rod with neutron transport code. OpenFOAM sends coolant density information and fuel temperature to the neutron transport code that sends back power information. A mapping function is used to average values when one node in one side corresponds to many nodes in the other side. Data is exchanged between codes by library calls. As the results of a fuel rod calculations progress, more complicated and processing demanding geometries will be simulated, aiming to the simulation of a real scale PWR fuel assembly

  14. Hierarchic modeling of heat exchanger thermal hydraulics

    International Nuclear Information System (INIS)

    Horvat, A.; Koncar, B.

    2002-01-01

    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 C d as a function of Reynolds number Re h . 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 C d are large, we believe that hierarchic modeling is an appropriate strategy for calculation of complex transport phenomena in heat exchanger geometries.(author)

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

    International Nuclear Information System (INIS)

    Tuomisto, Harri.

    1987-06-01

    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

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

    International Nuclear Information System (INIS)

    W.E. Lowry

    2001-01-01

    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

  17. Thermal-hydraulic experiments and analyses on cold moderator

    International Nuclear Information System (INIS)

    Aso, Tomokazu; Kaminaga, Masanori; Terada, Atsushi; Hino, Ryutaro

    2001-01-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Chellapandi, P.; Velusamy, K., E-mail: kvelu@igcar.gov.in

    2015-12-01

    Highlights: • We present challenges in thermal hydraulic design of sodium cooled fast reactors. • We present roadmap of Indian fast reactor program and innovative design concepts. • Analysis methodology for thermal striping and thermal stratification are highlighted. • Design solutions for gas entrainment are presented. • Experimental approaches for normal and post accident decay heat removal are highlighted. - Abstract: Pool type sodium cooled fast reactors pose several design challenges and among them, certain thermal hydraulics and structural mechanics issues are special. High frequency temperature fluctuations due to thermal striping, thermal stratifications and sodium free level fluctuations at the liquid–cover gas interfaces are to be investigated carefully to eliminate high cycle thermal fatigue of structures. Solutions to address the core thermal hydraulics call for high power computing. Innovative concepts and methods are developed to carry out plant dynamics and safety studies. Particularly, extensive numerical and experimental simulation techniques are needed for understanding and solving the gas entrainment mechanisms and its effects on core safety. Though decay heat removal through natural convection is achievable in a pool type SFR, demonstration of design solutions conceived in the reactor and performance of diverse systems under all operating conditions, especially over prolonged station blackout situations needs advanced CFD computations and should be validated by relatively large scale simulated experiments. These issues are addressed in this paper under five broad topics: special thermal hydraulic issues to be addressed in SFR, thermal hydraulic design and analysis, plant dynamics studies, safety studies and evolving thermal hydraulic studies for the future FBRs. The 500 MWe Prototype Fast Breeder Reactor (PFBR) is taken as the reference design for addressing the issues. Indian fast reactor programme is highlighted in the introduction

  19. Analyses of hydraulic performance of velocity caps

    DEFF Research Database (Denmark)

    Christensen, Erik Damgaard; Degn Eskesen, Mark Chr.; Buhrkall, Jeppe

    2014-01-01

    The hydraulic performance of a velocity cap has been investigated. Velocity caps are often used in connection with offshore intakes. CFD (computational fluid dynamics) examined the flow through the cap openings and further down into the intake pipes. This was combined with dimension analyses...

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

    International Nuclear Information System (INIS)

    Ville Lestinen; Timo Toppila

    2005-01-01

    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

  1. Validation of containment thermal hydraulic computer codes for VVER reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jiri Macek; Lubomir Denk [Nuclear Research Institute Rez plc Thermal-Hydraulic Analyses Department CZ 250 68 Husinec-Rez (Czech Republic)

    2005-07-01

    Full text of publication follows: The Czech Republic operates 4 VVER-440 units, two VVER-1000 units are being finalized (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-suppression system are modelled with COCOSYS 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. One of the important components of the VVER 440/213 NPP is its containment with pressure suppression system (bubble condenser). For safety analyses of this system, computer codes of the type MELCOR and COCOSYS are used in the Czech Republic. These codes were developed for containments of classic PWRs or BWRs. In order to apply these codes for VVER 440 systems, their validation on experimental facilities must be performed.The paper provides concise information on these activities of the NRI and its Thermal-hydraulics Department. The containment system of the VVER 440/213, its functions and approaches to solution of its safety is described with definition of acceptance criteria. A detailed example of the containment code validation on EREC Test facility (LOCA and MSLB) and the consequent utilisation of the results for a real NPP purposes is included. An approach to

  2. 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.

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

    International Nuclear Information System (INIS)

    Mur, J.; Meignin, J.C.

    1997-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  5. submitter Thermal, Hydraulic, and Electromagnetic Modeling of Superconducting Magnet Systems

    CERN Document Server

    Bottura, L

    2016-01-01

    Modeling techniques and tailored computational tools are becoming increasingly relevant to the design and analysis of large-scale superconducting magnet systems. Efficient and reliable tools are useful to provide an optimal forecast of the envelope of operating conditions and margins, which are difficult to test even when a prototype is available. This knowledge can be used to considerably reduce the design margins of the system, and thus the overall cost, or increase reliability during operation. An integrated analysis of a superconducting magnet system is, however, a complex matter, governed by very diverse physics. This paper reviews the wide spectrum of phenomena and provides an estimate of the time scales of thermal, hydraulic, and electromagnetic mechanisms affecting the performance of superconducting magnet systems. The analysis is useful to provide guidelines on how to divide the complex problem into building blocks that can be integrated in a design and analysis framework for a consistent multiphysic...

  6. Thermal hydraulic analysis of the encapsulated nuclear heat source

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, J.J.; Wade, D.C. [Argonne National Lab., IL (United States)

    2001-07-01

    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)

  7. Scaling in nuclear reactor system thermal-hydraulics

    International Nuclear Information System (INIS)

    D'Auria, F.; Galassi, G.M.

    2010-01-01

    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.

  8. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Mushtaq, A. [Isotope Production Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)], E-mail: mushtaqa@pinstech.org.pk; Iqbal, Massod; Bokhari, Ishtiaq Hussain; Mahmood, Tariq; Mahmood, Tayyab; Ahmad, Zahoor; Zaman, Qamar [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)

    2008-02-15

    Neutronic and thermal hydraulic analysis for the fission molybdenum-99 production at PARR-1 has been performed. Low enriched uranium foil (<20% {sup 235}U) 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 {sup 99}Mo/{sup 99m}Tc 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.

  10. Thermal-hydraulic design of the 200 MW NHR

    Energy Technology Data Exchange (ETDEWEB)

    Jincai, Li; Zuying, Gao; Baocheng, Xu; Junxiao, He [Institute of Nuclear Energy and Technology, Tsingua Univ., Beijing (China)

    1997-09-01

    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.

  11. Thermal-hydraulic methods in fast reactor safety

    International Nuclear Information System (INIS)

    Weber, D.P.; Briggs, L.L.

    1985-01-01

    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

  12. Design and thermal-hydraulic calculation for EAST PFCs' baking

    International Nuclear Information System (INIS)

    Wan Xiaogang; Yao Damao

    2006-01-01

    According to the vacuum requirements for fusion in a tokamak device, the authors adopted a kind of gas flow baking technique in EAST. This paper presented the sketch design for EAST PFCs' baking, selected the specifications for the working gas. Calculated the hydraulic and thermal conditions in PFCs under baking, and simulated the results. (authors)

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

    International Nuclear Information System (INIS)

    Lee, R.Y.

    1988-02-01

    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

  14. A review on the thermal hydraulic characteristics of the air-cooled

    Indian Academy of Sciences (India)

    In this paper, a review is presented on the experimental investigations and the numerical simulations performed to analyze the thermal-hydraulic performance of the air-cooled heat exchangers. The air-cooled heat exchangers mostly consist of the finned-tube bundles. The primary role of the extended surfaces (fins) is to ...

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

    Energy Technology Data Exchange (ETDEWEB)

    HEARD, F.J.

    1999-04-08

    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.

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

    International Nuclear Information System (INIS)

    Heard, F.J.

    1999-01-01

    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

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

    International Nuclear Information System (INIS)

    Vaiana, F.

    2009-11-01

    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)

  18. Status and subjects of thermal-hydraulic analysis for next-generation LWRs

    International Nuclear Information System (INIS)

    2000-03-01

    The status and subjects on thermal-hydraulic analysis for next-generation light water reactors (LWRs) with passive safety systems were surveyed through about 5 years until March 1999 by subcommittee on improvement of reactor thermal-hydraulic analysis codes under the nuclear code committee in Japan Atomic Energy Research Institute. Based on the survey results and discussion, the status and subjects on system analysis for various types of proposed reactor were summarized in 1998 and those on multidimensional two-phase flow analysis were also reviewed, since the multidimensional analysis was recognized as one of the most important subjects through the investigation on system analysis. In this report, the status and subjects for the following were summarized from the survey results and discussion in 1998 and 1999; (1) BWR neutronic/thermal-hydraulic coupled analysis, (2) Evaluation of passive safety system performance and (3) Gas-liquid two-phase flow analysis. The contents in this report are the forefront of thermal-hydraulic analysis for LWRs including test results from several large-scale facilities. We expect that the contents can offer a guideline to improve reactor thermal-hydraulic analysis codes in future. (author)

  19. 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.

  20. CCP Sensitivity Analysis by Variation of Thermal-Hydraulic Parameters of Wolsong-3, 4

    Energy Technology Data Exchange (ETDEWEB)

    You, Sung Chang [KHNP, Daejeon (Korea, Republic of)

    2016-10-15

    The PHWRs are tendency that ROPT(Regional Overpower Protection Trip) setpoint is decreased with reduction of CCP(Critical Channel Power) due to aging effects. For this reason, Wolsong unit 3 and 4 has been operated less than 100% power due to the result of ROPT setpoint evaluation. Typically CCP for ROPT evaluation is derived at 100% PHTS(Primary Heat Transport System) boundary conditions - inlet header temperature, header to header different pressure and outlet header pressure. Therefore boundary conditions at 100% power were estimated to calculate the thermal-hydraulic model at 100% power condition. Actually thermal-hydraulic boundary condition data for Wolsong-3 and 4 cannot be taken at 100% power condition at aged reactor condition. Therefore, to create a single-phase thermal-hydraulic model with 80% data, the validity of the model was confirmed at 93.8%(W3), 94.2%(W4, in the two-phase). And thermal-hydraulic boundary conditions at 100% power were calculated to use this model. For this reason, the sensitivities by varying thermal-hydraulic parameters for CCP calculation were evaluated for Wolsong unit 3 and 4. For confirming the uncertainties by variation PHTS model, sensitivity calculations were performed by varying of pressure tube roughness, orifice degradation factor and SG fouling factor, etc. In conclusion, sensitivity calculation results were very similar and the linearity was constant.

  1. 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.

  2. Thermal hydraulic analyses of two fusion reactor first wall/blanket concepts

    International Nuclear Information System (INIS)

    Misra, B.; Maroni, V.A.

    1978-01-01

    A comparative study has been made of the thermal hydraulic performance of two liquid lithium blanket concepts for tokamak-type reactors. In one concept lithium is circulated through 60-cm deep cylindrical modules oriented so that the module axis is parallel to the reactor minor radius. In the other concept helium carrying channels oriented parallel to the first wall are used to cool a 60-cm thick stagnant lithium blanket. Paralleling studies were carried out wherein the thermal and structural properties of the construction materials were based on those projected for either solution-annealed 316-stainless steel or vanadium-base alloys. The effects of limitations on allowable peak structural temperature, material strength, thermal stress, coolant inlet temperature, and pumping power/thermal power ratio were evaluated. Consequences to thermal hydraulic performance resulting from the presence of or absence of a divertor were also investigated

  3. Thermal hydraulic analyses of two fusion reactor first wall/blanket concepts

    International Nuclear Information System (INIS)

    Misra, B.; Maroni, V.A.

    1977-01-01

    A comparative study has been made of the thermal hydraulic performance of two liquid lithium blanket concepts for tokamak-type reactors. In one concept lithium is circulated through 60-cm deep cylindrical modules oriented so that the module axis is parallel to the reactor minor radius. In the other concept helium carrying channels oriented parallel to the first wall are used to cool a 60-cm thick stagnant lithium blanket. Paralleling studies were carried out wherein the thermal and structural properties of the construction materials were based on those projected for either solution-annealed 316-stainless steel or vanadium-base alloys. The effects of limitations on allowable peak structural temperature, material strength, thermal stress, coolant inlet temperature, and pumping power/thermal power ratio were evaluated. Consequences to thermal hydraulic performance resulting from the presence of or absence of a divertor were also investigated

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

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Park, Chun Kyeong; Yang, Seon Kyu; Chung, Chang Hwan; Chun, Shee Yeong; Song, Cheol Hwa; Chun, Hyeong Gil; Chang, Seok Kyu; Chung, Heung Joon; Won, Soon Yeon; Cho, Yeong Ro; Kim, Bok Deuk; Min, Kyeong Ho

    1994-07-01

    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)

  5. 78 FR 8202 - Meeting of the Joint ACRS Subcommittees on Thermal Hydraulic Phenomena and Materials, Metallurgy...

    Science.gov (United States)

    2013-02-05

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Joint ACRS Subcommittees on Thermal Hydraulic Phenomena and Materials, Metallurgy and Reactor Fuels; Notice of Meeting The Joint ACRS Subcommittees on Thermal Hydraulic Phenomena and Materials, Metallurgy and...

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

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jee Won; Chae, Kyung Myung; Choi, Hang Bok

    2000-07-01

    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.

  7. FX2-TH: a two-dimensional nuclear reactor kinetics code with thermal-hydraulic feedback

    International Nuclear Information System (INIS)

    Shober, R.A.; Daly, T.A.; Ferguson, D.R.

    1978-10-01

    FX2-TH is a two-dimensional, time-dependent nuclear reactor kinetics program with thermal and hydraulic feedback. The neutronics model used is multigroup neutron diffusion theory. The following geometry options are available: x, r, x-y, r-z, theta-r, and triangular. FX2-TH contains two basic thermal and hydraulic models: a simple adiabatic fuel temperature calculation, and a more detailed model consisting of an explicit representation of a fuel pin, gap, clad, and coolant. FX2-TH allows feedback effects from both fuel temperature (Doppler) and coolant temperature (density) changes. FX2-TH will calculate a consistent set of steady state conditions by iterating between the neutronics and thermal-hydraulics until convergence is reached. The time-dependent calculation is performed by the use of the improved quasistatic method. A disk editing capability is available. FX2-TH is operational on IBM system 360 or 370 computers and on the CDC 7600

  8. Primary system thermal hydraulics of future Indian fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Velusamy, K., E-mail: kvelu@igcar.gov.in [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Natesan, K.; Maity, Ram Kumar; Asokkumar, M.; Baskar, R. Arul; Rajendrakumar, M.; Sarathy, U. Partha; Selvaraj, P.; Chellapandi, P. [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Kumar, G. Senthil; Jebaraj, C. [AU-FRG Centre for CAD/CAM, Anna University, Chennai 600 025 (India)

    2015-12-01

    Highlights: • We present innovative design options proposed for future Indian fast reactor. • These options have been validated by extensive CFD simulations. • Hotspot factors in fuel subassembly are predicted by parallel CFD simulations. • Significant safety improvement in the thermal hydraulic design is quantified. - Abstract: As a follow-up to PFBR (Indian prototype fast breeder reactor), many FBRs of 500 MWe capacity are planned. The focus of these future FBRs is improved economy and enhanced safety. They are envisaged to have a twin-unit concept. Design and construction experiences gained from PFBR project have provided motivation to achieve an optimized design for future FBRs with significant design changes for many critical components. Some of the design changes include, (i) provision of four primary pipes per primary sodium pump, (ii) inner vessel with single torus lower part, (iii) dome shape 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. This paper covers thermal hydraulic design validation of the chosen options with respect to hot and cold pool thermal hydraulics, flow requirement for main vessel cooling, inner vessel temperature distribution, safety analysis of primary pipe rupture event, adequacy of decay heat removal capacity by natural convection cooling, cold pool transient thermal loads and thermal management of top shield and reactor vault.

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

    International Nuclear Information System (INIS)

    Hennion, F.

    1987-10-01

    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 [fr

  10. TRAC-B thermal-hydraulic analysis of the Black Fox boiling water reactor

    International Nuclear Information System (INIS)

    Martin, R.P.

    1993-05-01

    Thermal-hydraulic analyses of six hypothetical accident scenarios for the General Electric Black Fox Nuclear Project boiling water reactor were performed using the TRAC-BF1 computer code. This work is sponsored by the US Nuclear Regulatory Commission and is being done in conjunction with future analysis work at the US Nuclear Regulatory Commission Technical Training Center in Chattanooga, Tennessee. These accident scenarios were chosen to assess and benchmark the thermal-hydraulic capabilities of the Black Fox Nuclear Project simulator at the Technical Training Center to model abnormal transient conditions

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

    Energy Technology Data Exchange (ETDEWEB)

    Basehore, K.L.; Todreas, N.E.

    1980-08-01

    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.

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

    International Nuclear Information System (INIS)

    Basehore, K.L.; Todreas, N.E.

    1980-08-01

    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

  13. Core Thermal-Hydraulic Conceptual Design for the Advanced SFR Design Concepts

    International Nuclear Information System (INIS)

    Cho, Chung Ho; Chang, Jin Wook; Yoo, Jae Woon; Song, Hoon; Choi, Sun Rock; Park, Won Seok; Kim, Sang Ji

    2010-01-01

    The Korea Atomic Energy Research Institute (KAERI) has developed the advanced SFR design concepts from 2007 to 2009 under the National longterm Nuclear R and D Program. Two types of core designs, 1,200 MWe breakeven and 600 MWe TRU burner core have been proposed and evaluated whether they meet the design requirements for the Gen IV technology goals of sustainability, safety and reliability, economics, proliferation resistance, and physical protection. In generally, the core thermal hydraulic design is performed during the conceptual design phase to efficiently extract the core thermal power by distributing the appropriate sodium coolant flow according to the power of each assembly because the conventional SFR core is composed of hundreds of ducted assemblies with hundreds of fuel rods. In carrying out the thermal and hydraulic design, special attention has to be paid to several performance parameters in order to assure proper performance and safety of fuel and core; the coolant boiling, fuel melting, structural integrity of the components, fuel-cladding eutectic melting, etc. The overall conceptual design procedure for core thermal and hydraulic conceptual design, i.e., flow grouping and peak pin temperature calculations, pressure drop calculations, steady-state and detailed sub-channel analysis is shown Figure 1. In the conceptual design phase, results of core thermal-hydraulic design for advanced design concepts, the core flow grouping, peak pin cladding mid-wall temperature, and pressure drop calculations, are summarized in this study

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

    International Nuclear Information System (INIS)

    Mesquita, Amir Z.; Costa, Antonio C.L.; Ladeira, Luiz C.D.; Rezende, Hugo C.; Palma, Daniel A.P.

    2015-01-01

    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)

  15. First wall thermal hydraulic models for fusion blankets

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1980-01-01

    Subject to normal and off-normal reactor conditions, thermal hydraulic models of first walls, e.g., a thermal mass barrier, a tubular shield, and a radiating liner are reviewed. Under normal operation the plasma behaves as expected in a predicted way for transient and steady-state conditions. The most severe thermal loading on the first wall occurs when the plasma becomes unstable and dumps its energy on the wall in a very short period of time (milliseconds). Depending on the plasma dump time and area over which the energy is deposited may result in melting of the first wall surface, and if the temperature is high enough, vaporization

  16. Project W-320 thermal hydraulic model benchmarking and baselining

    International Nuclear Information System (INIS)

    Sathyanarayana, K.

    1998-01-01

    Project W-320 will be retrieving waste from Tank 241-C-106 and transferring the waste to Tank 241-AY-102. Waste in both tanks must be maintained below applicable thermal limits during and following the waste transfer. Thermal hydraulic process control models will be used for process control of the thermal limits. This report documents the process control models and presents a benchmarking of the models with data from Tanks 241-C-106 and 241-AY-102. Revision 1 of this report will provide a baselining of the models in preparation for the initiation of sluicing

  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. Thermal hydraulics analysis of the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Dean, E-mail: Dean_Wang@uml.edu [University of Massachusetts Lowell, One University Avenue, Lowell, MA 01854 (United States); Yoder, Graydon L.; Pointer, David W.; Holcomb, David E. [Oak Ridge National Laboratory, 1 Bethel Valley RD #6167, Oak Ridge, TN 37831 (United States)

    2015-12-01

    Highlights: • The TRACE AHTR model was developed and used to define and size the DRACS and the PHX. • A LOFF transient was simulated to evaluate the reactor performance during the transient. • Some recommendations for modifying FHR reactor system component designs are discussed. - Abstract: The Advanced High Temperature Reactor (AHTR) is a liquid salt-cooled nuclear reactor design concept, featuring low-pressure molten fluoride salt coolant, a carbon composite fuel form with embedded coated particle fuel, passively triggered negative reactivity insertion mechanisms, and fully passive decay heat rejection. This paper describes an AHTR system model developed using the Nuclear Regulatory Commission (NRC) thermal hydraulic transient code TRAC/RELAP Advanced Computational Engine (TRACE). The TRACE model includes all of the primary components: the core, downcomer, hot legs, cold legs, pumps, direct reactor auxiliary cooling system (DRACS), the primary heat exchangers (PHXs), etc. The TRACE model was used to help define and size systems such as the DRACS and the PHX. A loss of flow transient was also simulated to evaluate the performance of the reactor during an anticipated transient event. Some initial recommendations for modifying system component designs are also discussed. The TRACE model will be used as the basis for developing more detailed designs and ultimately will be used to perform transient safety analysis for the reactor.

  19. Thermal-hydraulics associated with nuclear education and research

    International Nuclear Information System (INIS)

    Yokobori, Seiichi

    2011-01-01

    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)

  20. Thermal-hydraulic characteristics of double flat core HCLWR

    International Nuclear Information System (INIS)

    Sugimoto, Jun; Iwamura, Takamichi; Okubo, Tsutomu; Murao, Yoshio

    1989-02-01

    A thermal-hydraulic characteristics of double flat core high conversion light water reactor (HCLWR) is described. The concept of flat core proposed by Ishiguro et al. is to achieve negative void reactivity coefficient in tight lattice core, and at the same time, high conversion ratio and high burnup can be obtainable. The proposed double flat core HCLWR, based on these physical advantages and the consideration of safety assurance, aims at efficient use of the pressure vessel space to produce comparable thermal output as current 3-loop PWRs. The present work revealed the following items concerning the thermalhydraulic feasibility of the double flat core HCLWR: (1) Main thermal-hydraulic parameters of the plant can be almost the same as current PWRs, showing the use of PWR standard components without major modifications except in core region. (2) Heat removal from the fuel rod in a steady operational condition has enough margin to the critical heat flux (CHF) limit, which is evaluated with the existing CHF correlations. (3) The calculation by REFLA code shows that the maximum cladding temperature in LOCA-reflood is estimated to be far lower than the licensing criteria. It is therefore considered that the proposed double flat core HCLWR is feasible from the point of thermal-hydraulics. Since the available data base has certain applicational limit to the very short core as the present double flat core HCLWR, further detailed assessment is required. (author)

  1. 11. international topical meeting on nuclear reactor thermal-hydraulics (NURETH-11)

    International Nuclear Information System (INIS)

    Lemonnier, H.

    2005-01-01

    ; aerosol transport, deposition and re-entrainment; steam generators thermal-hydraulics; system codes development and assessment; uncertainties analysis; diffuse interface methods and interface tracking methods; C - severe accidents and fires: molten core natural convection and physico-chemical phenomena, modeling and experiments; fuel coolant interaction, modeling and experiments; debris bed cooling; combustion and fires, modeling and experiments; molten corium concrete interaction; D - advanced code developments: fast transient modelling and experiments; multidimensional single-phase or two-phase flow and heat transfer modeling; neutronics and thermal-hydraulics coupling; fluid and structures mechanical interactions; coupled thermal-hydraulics of fluids and structures; thermal-hydraulic dependent corrosion and ablation; E - operation and safety of existing reactors: instabilities and nonlinear dynamics; NPP transients and accidents analysis; RBMK and VVER safety analysis, including the OECD benchmark; F - experimental thermal-hydraulics: boiling heat transfer; CHF and post-CHF heat transfer; condensation heat transfer; integral testing; vibrations, wear and thermal fatigue phenomena; fuel design and performance; G - advanced reactors thermal-hydraulics (gen IV, INPRO, fusion, hydrogen production): accelerator driven reactors; advanced pressurized water reactors thermal-hydraulics; gas cooled fast reactors; gas cooled high temperature reactors; lead and lead-bismuth cooled reactors; future and existing sodium cooled reactors; molten salt reactors; H - waste management thermal-hydraulics: thermal-hydraulics problems related to waste processing and storage; I - thermal-hydraulics of non electricity generating nuclear equipment: sono-fusion (cavitation induced bubble fusion; hydrogen producing nuclear reactors

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

    International Nuclear Information System (INIS)

    Determan, J.C.; Hendrix, C.E.

    1992-12-01

    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

  3. Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly

    International Nuclear Information System (INIS)

    Liu, X.J.; Cheng, X.

    2009-01-01

    A new fuel assembly design for a thermal supercritical water cooled reactor (SCWR) core is proposed. Compared to the existing fuel assemblies, the present fuel assembly has two-rows of fuel rods between the moderator channels, to achieve a more uniform moderation for all fuel rod cells, and subsequently, a more uniform radial power distribution. In addition, a neutron-kinetics/thermal-hydraulics coupling method is developed, to analyze the neutron-physical and thermal-hydraulic behavior of the fuel assembly designs. This coupling method is based on the sub-channel analysis code COBRA-IV for thermal-hydraulics and the neutron-kinetics code SKETCH-N for neutron-physics. Both the COBRA-IV code and the SKETCH-N code are accordingly modified. An interface is established for the data transfer between these two codes. This coupling method is applied to both the one-row fuel assemblies (previous design) and the two-row fuel assemblies (present design). The performance of the two types of fuel assemblies is compared. The results show clearly that the two-row fuel assembly has more favorable neutron-physical and thermal-hydraulic characteristics than the one-row fuel assembly. The effect of various parameters on the fuel assembly performance is discussed. The coupling method is proven to be well suitable for further applications to SCWR fuel assembly design analysis

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

    International Nuclear Information System (INIS)

    Kim, Hee-Kyung; Kim, Soo Hyoung; Chung, Young-Jong; Kim, Hyeon-Soo

    2013-01-01

    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

  5. 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.

  6. THERMAL HYDRAULIC ANALYSIS OF FIRE DIVERTOR

    International Nuclear Information System (INIS)

    C.B. bAXI; M.A. ULRICKSON; D.E. DRIMEYER; P. HEITZENROEDER

    2000-01-01

    The Fusion Ignition Research Experiment (FIRE) is being designed as a next step in the US magnetic fusion program. The FIRE tokamak has a major radius of 2 m, a minor radius of 0.525 m, and liquid nitrogen cooled copper coils. The aim is to produce a pulse length of 20 s with a plasma current of 6.6 MA and with alpha dominated heating. The outer divertor and baffle of FIRE are water cooled. The worst thermal condition for the outer divertor and baffle is the baseline D-T operating mode (10 T, 6.6 MA, 20 s) with a plasma exhaust power of 67 MW and a peak heat flux of 20 MW/m 2 . A swirl tape (ST) heat transfer enhancement method is used in the outer divertor cooling channels to increase the heat transfer coefficient and the critical heat flux (CHF). The plasma-facing surface consists of tungsten brush. The finite element (FE) analysis shows that for an inlet water temperature of 30 C, inlet pressure of 1.5 MPa and a flow velocity of 10 m/s, the incident critical heat flux is greater than 30 MW/m 2 . The peak copper temperature is 490 C, peak tungsten temperature is 1560 C, and the pressure drop is less than 0.5 MPa. All these results fulfill the design requirements

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

    International Nuclear Information System (INIS)

    Jimenez, J.; Herrero, J. J.; Cuervo, D.; Aragones, J. M.

    2010-10-01

    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)

  8. Coupled 3D neutronics/thermal hydraulics modeling of the SAFARI-1 MTR

    International Nuclear Information System (INIS)

    Rosenkrantz, Adam; Avramova, Maria; Ivanov, Kostadin; Prinsloo, Rian; Botes, Danniëll; Elsakhawy, Khalid

    2014-01-01

    Highlights: • Development of 3D coupled neutronics/thermal–hydraulic model of SAFARI-1. • Verification of 3D steady-state NEM based neutronics model for SAFARI-1. • Verification of 3D COBRA-TF based thermal–hydraulic model of SAFARI-1. • Quantification of the effect of correct modeling of thermal–hydraulic feedback. - Abstract: The purpose of this study was to develop a coupled accurate multi-physics model of the SAFARI-1 Material Testing Reactor (MTR), a facility that is used for both research and the production of medical isotopes. The model was developed as part of the SAFARI-1 benchmarking project as a cooperative effort between the Pennsylvania State University (PSU) and the South African Nuclear Energy Corporation (Necsa). It was created using a multi-physics coupling of state of the art nuclear reactor simulation tools, consisting of a neutronics code and a thermal hydraulics code. The neutronics tool used was the PSU code NEM, and the results from this component were verified using the Necsa neutronics code OSCAR-4, which is utilized for SAFARI-1 core design and fuel management. On average, the multiplication factors of the neutronics models agreed to within 5 pcm and the radial assembly-averaged powers agreed to within 0.2%. The thermal hydraulics tool used was the PSU version of COBRA-TF (CTF) sub-channel code, and the results of this component were verified against another thermal hydraulics code, the RELAP5-3D system code, used at Necsa for thermal–hydraulics analysis of SAFARI-1. Although only assembly-averaged results from RELAP5-3D were available, they fell within the range of values for the corresponding assemblies in the comprehensive CTF solution. This comparison allows for the first time to perform a quantification of steady-state errors for a low-powered MTR with an advanced thermal–hydraulic code such as CTF on a per-channel basis as compared to simpler and coarser-mesh RELAP5-3D modeling. Additionally, a new cross section

  9. Optimised Iteration in Coupled Monte Carlo - Thermal-Hydraulics Calculations

    Science.gov (United States)

    Hoogenboom, J. Eduard; Dufek, Jan

    2014-06-01

    This paper describes an optimised iteration scheme for the number of neutron histories and the relaxation factor in successive iterations of coupled Monte Carlo and thermal-hydraulic reactor calculations based on the stochastic iteration method. The scheme results in an increasing number of neutron histories for the Monte Carlo calculation in successive iteration steps and a decreasing relaxation factor for the spatial power distribution to be used as input to the thermal-hydraulics calculation. The theoretical basis is discussed in detail and practical consequences of the scheme are shown, among which a nearly linear increase per iteration of the number of cycles in the Monte Carlo calculation. The scheme is demonstrated for a full PWR type fuel assembly. Results are shown for the axial power distribution during several iteration steps. A few alternative iteration method are also tested and it is concluded that the presented iteration method is near optimal.

  10. Thermal-hydraulic modeling of porous bed reactors

    International Nuclear Information System (INIS)

    Araj, K.J.; Nourbakhsh, H.P.

    1987-01-01

    Optimum design of nuclear reactor core requires an iterative approach between the thermal-hydraulic, neutronic and operational analysis. This paper concentrates on the thermal-hydraulic behavior of a hydrogen cooled, small particle bed reactor (PBR). The PBR core, modeled here, consists of a hexagonal array of fuel elements embedded in a moderator matrix. The fuel elements are annular packed beds of fuel particles held between two porous cylindrical frits. These particles, 500 to 600 μm in diameter, have a uranium carbide core, which is coated by two layers of graphite and an outer coating of zirconium carbide. Coolant flow, radially inward, from the cold frit through the packed bed and hot frit and axially out the channel, formed by the hot frit, to a common plenum. 5 refs., 1 fig., 2 tabs

  11. 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.

  12. Simulation of Thermal Hydraulic at Supercritical Pressures with APROS

    Energy Technology Data Exchange (ETDEWEB)

    Kurki, Joona [VTT Technical Research Centre of Finland, P.O. Box 1000, FI02044 VTT (Finland)

    2008-07-01

    The proposed concepts for the fourth generation of nuclear reactors include a reactor operating with water at thermodynamically supercritical state, the Supercritical Water Reactor (SCWR). For the design and safety demonstrations of such a reactor, the possibility to accurately simulate the thermal hydraulics of the supercritical coolant is an absolute prerequisite. For this purpose, the one-dimensional two-phase thermal hydraulics solution of APROS process simulation software was developed to function at the supercritical pressure region. Software modifications included the redefinition of some parameters that have physical significance only at the subcritical pressures, improvement of the steam tables, and addition of heat transfer and friction correlations suitable for the supercritical pressure region. (author)

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

    International Nuclear Information System (INIS)

    Soldermann, R.; Poizat, F.; Sekri, A.; Faydide, B.; Dumas, J.M.

    1997-01-01

    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

  14. A study on the thermal hydraulics in rod bundles

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Yang, Sun Kyu

    1989-03-01

    In order to improve the thermal hydraulic characteristics of the nuclear reactor core, it is necessary to obtain better understanding of the coolant flow and the enthalpy distribution in complex rod bundle geometries. The purpose of this report is to obtain a comprehensive survey on the thermal hydraulic in rod bundles from both experimental and numerical point of view. From references on experimental study, measurement methods and results of the flow velocity and the pressure drop in the subchannels of rod bundles are expressed. The microscopic flow characteristics of the subchannels and spacer grid effect on the flow structure are described. Physical phenomena and measurement methods of the secondary flow are also described. From references on the numerical study, general numerical methods are expressed. Numerical studies on the laminar flow and turbulent flow such as 1-equation and 2-equation model are reviewed.(Author)

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

    International Nuclear Information System (INIS)

    Denny, V.E.; Wassel, A.T.; Issacci, F.; Pal Kalra, S.

    2004-01-01

    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)

  16. State of the art of thermal-hydraulics of BWRs

    International Nuclear Information System (INIS)

    Rouhani, Z.

    1980-10-01

    The present report is a summary review of the developments in the field of thermal-hydraulics of Boiling Water Reactors. It covers briefly the development of BWR systems, including some comparison of the main features of the modern BWRs that are marketed by different vendors. The analytical aspects of BWR are also covered briefly with some remarks on the problem areas and limitations in this field. (author)

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

    International Nuclear Information System (INIS)

    Queral, C.; Coriso, M.; Garcia Sedano, P. J.; Ruiz, J. A.; Posada, J. M.; Jimenez Varas, G.; Sol, I.; Herranz, L. E.

    2011-01-01

    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)

  18. 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.

  19. Thermal-hydraulic interfacing code modules for CANDU reactors

    International Nuclear Information System (INIS)

    Liu, W.S.; Gold, M.; Sills, H.

    1997-01-01

    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

  20. Computerized hydraulic and thermal laboratory investigations

    International Nuclear Information System (INIS)

    Lepetit, J.P.; Manoha, B.

    1981-02-01

    The paper presents the principles and capabilities of a computer controlled tank performed by the L.N.H. to assume an automatic reproduction of non steady currents. It gives a description of the tank and its equipment for model control, measurement and treatment of data; it shows an example of results obtained on the site of the Flamanville nuclear plant [fr

  1. 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.

  2. Views on the future of thermal hydraulic modeling

    International Nuclear Information System (INIS)

    Ishii, M.

    1997-01-01

    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

  3. Development of design technology on thermal-hydraulic performance in tight-lattice rod bundle. 4. Large paralleled simulation by the advanced two-fluid model code

    International Nuclear Information System (INIS)

    Misawa, Takeharu; Yoshida, Hiroyuki; Akimoto, Hajime

    2008-01-01

    In Japan Atomic Energy Agency (JAEA), the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been developed. For thermal design of FLWR, it is necessary to develop analytical method to predict boiling transition of FLWR. Japan Atomic Energy Agency (JAEA) has been developing three-dimensional two-fluid model analysis code ACE-3D, which adopts boundary fitted coordinate system to simulate complex shape channel flow. In this paper, as a part of development of ACE-3D to apply to rod bundle analysis, introduction of parallelization to ACE-3D and assessments of ACE-3D are shown. In analysis of large-scale domain such as a rod bundle, even two-fluid model requires large number of computational cost, which exceeds upper limit of memory amount of 1 CPU. Therefore, parallelization was introduced to ACE-3D to divide data amount for analysis of large-scale domain among large number of CPUs, and it is confirmed that analysis of large-scale domain such as a rod bundle can be performed by parallel computation with keeping parallel computation performance even using large number of CPUs. ACE-3D adopts two-phase flow models, some of which are dependent upon channel geometry. Therefore, analyses in the domains, which simulate individual subchannel and 37 rod bundle, are performed, and compared with experiments. It is confirmed that the results obtained by both analyses using ACE-3D show agreement with past experimental result qualitatively. (author)

  4. On-Line Core Thermal-Hydraulic Model Improvement

    International Nuclear Information System (INIS)

    In, Wang Kee; Chun, Tae Hyun; Oh, Dong Seok; Shin, Chang Hwan; Hwang, Dae Hyun; Seo, Kyung Won

    2007-02-01

    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

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

    International Nuclear Information System (INIS)

    Cheng, X.; Yu, Y.Q.

    2009-01-01

    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.

  6. On-Line Core Thermal-Hydraulic Model Improvement

    Energy Technology Data Exchange (ETDEWEB)

    In, Wang Kee; Chun, Tae Hyun; Oh, Dong Seok; Shin, Chang Hwan; Hwang, Dae Hyun; Seo, Kyung Won

    2007-02-15

    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.

  7. Thermal-hydraulic analysis of LTS cables for the DEMO TF coil using simplified models

    Directory of Open Access Journals (Sweden)

    Lewandowska Monika

    2017-03-01

    Full Text Available The conceptual design activities for the DEMOnstration reactor (DEMO – the prototype fusion power plant – are conducted in Europe by the EUROfusion Consortium. In 2015, three design concepts of the DEMO toroidal field (TF coil were proposed by Swiss Plasma Center (EPFL-SPC, PSI Villigen, Italian National Agency for New Technologies (ENEA Frascati, and Atomic Energy and Alternative Energies Commission (CEA Cadarache. The proposed conductor designs were subjected to complete mechanical, electromagnetic, and thermal-hydraulic analyses. The present study is focused on the thermal-hydraulic analysis of the candidate conductor designs using simplified models. It includes (a hydraulic analysis, (b heat removal analysis, and (c assessment of the maximum temperature and the maximum pressure in each conductor during quench. The performed analysis, aimed at verification whether the proposed design concepts fulfil the established acceptance criteria, provides the information for further improvements of the coil and conductors design.

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

    International Nuclear Information System (INIS)

    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

  9. 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.

  10. Thermal hydraulic simulation of moderator heat exchanger

    International Nuclear Information System (INIS)

    Anil Lal, S.; Rajakumar, A.; Vaidyanathan, G.; Srinivasan, R.; Chetal, S.C.

    1993-01-01

    Pressurized heavy water reactors form the majority in the first stage of India's nuclear power programme. Heavy water is both moderator and primary coolant. The heat generated in the moderator due to neutron moderation and capture has to be removed in moderator heat exchangers. It has been desired to improve the performance characteristics of moderator heat exchangers, whereby moderator would enter the calandria vessel at a low temperature and would enable higher power of operation for the same limiting temperature of moderator in the calandria. Results of studies carried out using a three dimensional computer code for various operating options are given. Using these velocities the heat exchangers have been analysed for flow induced vibrations. 7 refs., 6 figs., 6 tabs

  11. Interaction between thermal/hydraulics, human factors and system analysis for assessing feed and bleed risk benefits

    International Nuclear Information System (INIS)

    Lanore, J.M.; Caron, J.L.

    1987-11-01

    For probabilistic analysis of accident sequences, thermal/hydraulics, human factors and systems operation problems are frequently closely interrelated. This presentation will discuss a typical example which illustrates this interrelation: total loss of feedwater flow. It will present thermal/hydraulic analysises performed, how the T/H analysises are related to human factors and systems operation, and how, based on this, the failure probability of the feed and bleed cooling mode was evaluated

  12. A THERMAL-HYDRAULIC SYSTEM FOR THE CONVERSION AND THE STORAGE OF ENERGY

    Directory of Open Access Journals (Sweden)

    MITRAN Tudor

    2016-05-01

    Full Text Available The paper proposes the concept design of a thermal-hydraulic system that converts the thermal energy (from the geothermal water, from the cooling water of power equipment, from exhaust gasses, and so. in hydrostatic energy, that is stored in a hydraulic accumulator. The hydraulic energy can be converted into electrical energy when needed.

  13. An improved thermal-hydraulic modeling of the Jules Horowitz Reactor using the CATHARE2 system code

    Energy Technology Data Exchange (ETDEWEB)

    Pegonen, R., E-mail: pegonen@kth.se [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden); Bourdon, S.; Gonnier, C. [CEA, DEN, DER, SRJH, CEA Cadarache, 13108 Saint-Paul-lez-Durance Cedex (France); Anglart, H. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden)

    2017-01-15

    Highlights: • An improved thermal-hydraulic modeling of the JHR reactor is described. • Thermal-hydraulics of the JHR is analyzed during loss of flow accident. • The heat exchanger approach gives more realistic and less conservative results. - Abstract: The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support current and future nuclear reactor designs. The reactor is under construction at the CEA Cadarache research center in southern France and is expected to achieve first criticality at the end of this decade. This paper presents an improved thermal-hydraulic modeling of the reactor using solely CATHARE2 system code. Up to now, the CATHARE2 code was simulating the full reactor with a simplified approach for the core and the boundary conditions were transferred into the three-dimensional FLICA4 core simulation. A new more realistic methodology is utilized to analyze the thermal-hydraulic simulation of the reactor during a loss of flow accident.

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

    International Nuclear Information System (INIS)

    Lightston, M.F.; Rock, R.

    1996-01-01

    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)

  15. Thermal hydraulic test of advanced fuel bundle with spectral shift rod (SSR) for BWR. Effect of thermal hydraulic parameters on steady state characteristics

    International Nuclear Information System (INIS)

    Kondo, Takao; Kitou, Kazuaki; Chaki, Masao; Ohga, Yukiharu; Makigami, Takeshi

    2011-01-01

    Japanese national project of next generation light water reactor (LWR) development started in 2008. Under this project, spectral shift rod (SSR) is being developed. SSR, which replaces conventional water rod (WR) of boiling water reactor (BWR) fuel bundle, was invented to enhance the BWR's merit, spectral shift effect for uranium saving. In SSR, water boils by neutron and gamma-ray direct heating and water level is formed as a boundary of the upper steam region and the lower water region. This SSR water level can be controlled by core flow rate, which amplifies the change of average core void fraction, resulting in the amplified spectral shift effect. This paper presents the steady state test results of the base geometry case in SSR thermal hydraulic test, which was conducted under the national project of next generation LWR. In the test, thermal hydraulic parameters, such as flow rate, pressure, inlet subcooling and heater rod power are changed to evaluate these effects on SSR water level and other SSR characteristics. In the test results, SSR water level rose as flow rate rose, which showed controllability of SSR water level by flow rate. The sensitivities of other thermal hydraulic parameters on SSR water level were also evaluated. The obtained data of parameter's sensitivities is various enough for the further analytical evaluation. The fluctuation of SSR water level was also measured to be small enough. As a result, it was confirmed that SSR's steady state performance was as planned and that SSR design concept is feasible. (author)

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

    International Nuclear Information System (INIS)

    Guelfi, A.; Boucker, M.; Herard, J.M.; Peturaud, P.; Bestion, D.; Boudier, P.; Hervieu, E.; Fillion, P.; Grandotto, M.

    2007-01-01

    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

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

    International Nuclear Information System (INIS)

    Pan Jie; Yang Dong; Chen Gongming; Zhou Xu; Bi Qincheng

    2012-01-01

    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.

  18. Alternative solution algorithm for coupled thermal-hydraulic problems

    International Nuclear Information System (INIS)

    Farnsworth, D.A.; Rice, J.G.

    1986-01-01

    A thermal-hydraulic system involves flow of a fluid for which a combined solution of the continuity, momentum, and energy equations is required. When the solutions of the energy and momentum fields are dependent on each other, the system is said to be thermally coupled. A common problem encountered in the numerical solution of strongly coupled thermal-hydraulic problems is a very slow rate of convergence or a complete lack of convergence. Many times this degradation in convergence is due to the lack of true coupling between the energy and momentum fields during the iteration process. In the most widely used solution algorithms - such as the SIMPLE algorithm and its many variants - a sequential solution technique is required. That is, the solution process alternates between the flow and energy fields until a converged solution is obtained. This approach allows only implicit energy-momentum coupling. To improve the convergence rate for strongly coupled problems, a practical solution algorithm that can accommodate true energy-momentum coupling terms was developed. A complete simultaneous (versus sequential) solution of the governing conservation equations utilizing a line-by-line solution was developed and direct coupling terms between the momentum and energy fields were added utilizing a modified Newton-Raphson technique

  19. 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.

  20. Thermal-hydraulic analysis and design improvement for coolant channel of ITER shield block

    International Nuclear Information System (INIS)

    Zhao Ling; Li Huaqi; Zheng Jiantao; Yi Jingwei; Kang Weishan; Chen Jiming

    2013-01-01

    As an important part for ITER, shield block is used to shield the neutron heat. The structure design of shield block, especially the inner coolant channel design will influence its cooling effect and safety significantly. In this study, the thermal-hydraulic analysis for shield block has been performed by the computational fluid dynamics software, some optimization suggestions have been proposed and thermal-hydraulic characteristics of the improved model has been analyzed again. The analysis results for improved model show that pressure drop through flow path near the inlet and outlet region of the shield block has been reduced, and the total pressure drop in cooling path has been reduced too; the uniformity of the mass flowrate distribution and the velocity distribution have been improved in main cooling branches; the local highest temperature of solid domain reduced considerably, which could avoid thermal stress becoming too large because of coolant effect unevenly. (authors)

  1. Thermal-hydraulic tests on net divertor targets using swirl tubes

    International Nuclear Information System (INIS)

    Schlosser, J.; Chappuis, P.; Deschamps, P.; Massmann, P.; Falter, H.D.; Deschamps, G.H.

    1991-01-01

    Thermal-hydraulic tests have been carried out in collaboration between NET, CEA Cadarache and JET in order to find a cooling method capable of removing the high heat fluxes expected for the NET/ITER divertor. The goal was to evaluate by experiments the critical heat flux (CHF) and heat transfer in the subcooled boiling regime using twisted tapes as turbulence promoters and testing them under relevant thermal-hydraulic conditions. The CEA 200 kW Electron Beam (EB) facility and the 10 MW JET Neutral Beam (NB) test bed have been used to heat up the NET relevant test sections (TS) consisting of rectangular copper elements with circular internal channels. The TS have been exposed to the electron or ion beams under normal incidence. This paper reports the results of the experiments and of thermal analyses performed in support of the tests. The experimental CHF values have been benchmarked with the Tong-75 correlation

  2. Applications for coupled core neutronics and thermal-hydraulic models

    International Nuclear Information System (INIS)

    Eller, J.

    1996-01-01

    The unprecedented increases in computing capacity that have occurred during the last decade have affected our sciences, and thus our lives, to an extent that is difficult to overstate. All indications are that this trend will continue for years to come. Nuclear reactor systems analysis is one of many areas of engineering that has changed dramatically as a result of this evolution. Our ability to model the various mechanical and physical systems in greater and greater detail has allowed significant improvements in operational efficiency in spite of increasing regulatory requirements. Many of these efficiencies result from the use of more complex and geometrically detailed computer modeling, which is used to justify a reduction or elimination of some of the conservatisms required by earlier, less sophisticated analyses. And more recently, as our industries open-quotes downsize,close quotes efforts are being made to find ways to use the ever-increasing computing capacity to design systems that accomplish more work, in less time, and with fewer people. The balance of this paper discusses some of the visions that Duke Power Company feels would most benefit their particular methodologies. One of the concepts receiving a lot of attention involves an automated coupling of a thermal-hydraulic plant systems analysis model to a three-dimensional core neutronics program. The thermal-hydraulic analysis of several postulated system transients incorporates large conservatisms because of limited ability to model complex time-dependent asymmetric heat sources in adequate geometric detail. For these transients, the core behavior is closely coupled with the thermal-hydraulic behavior of the total plant system and vice versa. Steam-line break, uncontrolled rod withdrawal, and rod drop anayses are likely to benefit most from this type of linked process

  3. 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

  4. Thermal hydraulic behavior evaluation of tank A-101

    International Nuclear Information System (INIS)

    Ogden, D.M.

    1996-01-01

    This report describes a new evaluation conducted to help understand the thermal-hydraulic behavior of tank A-101. Prior analysis of temperature data indicated that the dome space and upper waste layer was slowly increasing in temperature increases are due to increasing ambient temperatures and termination of forced ventilation. However, this analysis also indicates that other dome cooling processes are slowly decreasing, or some slow increase in heating is occurring at the waste surface. Dome temperatures are not decreasing at the rate expected as a forced ventilation termination effects are accounted for

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

    International Nuclear Information System (INIS)

    Silva Galetti, M.R. da.

    1984-01-01

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

  6. Thermal-hydraulic considerations for particle bed reactors

    Science.gov (United States)

    Benenati, R.; Araj, K. J.; Horn, F.

    In the design of particle bed reactor (PBR) cores, consideration must be given to the gas coolant channels and their configuration. Neutronics analysis provides the relative volume fractions of the component materials, but these must be arranged in such a manner as to allow proper cooling of all components by the gas flow at relatively low pressure drops. The thermal hydraulic aspects of this problem are addressed. A description of the computer model used in the analysis of the steady state condition is also included. Blowdown tests on hot particle bed fuel elements were carried out and are described.

  7. Thermal hydraulic behavior of SCWR sliding pressure startup

    International Nuclear Information System (INIS)

    Fu Shengwei; Zhou Chong; Xu Zhihong; Yang Yanhua

    2011-01-01

    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)

  8. Study of thermal-hydraulic analyses with CIP method

    International Nuclear Information System (INIS)

    Doi, Yoshihiro

    1996-09-01

    New type of numerical scheme CIP has been proposed for solving hyperbolic type equations and the CIP is focused on as a less numerical diffusive scheme. C-CUP method with the CIP scheme is adopted to numerical simulations that treat compressible and incompressible fluids, phase change phenomena and Mixture fluids. To evaluate applicabilities of the CIP scheme and C-CUP method for thermal hydraulic analyses related to Fast Breeder Reactors (FBRs), the scheme and the method were reviewed. Feature of the CIP scheme and procedure of the C-CUP method were presented. The CIP scheme is used to solve linear hyperbolic type equations for advection term in basic equations of fluids. Key issues of the scheme is that profile between grid points is described to solve the equation by cubic polynomial and spatial derivatives of the polynomial. The scheme can capture steep change of solution and suppress numerical error. In the C-CUP method, the basic equations of fluids are divided into advection terms and the other terms. The advection terms is solved with CIP scheme and the other terms is solved with difference method. The C-CUP method is robust for numerical instability, but mass of fluid will be in unfair preservation with nonconservative equations for fluids. Numerical analyses with the CIP scheme and the C-CUP method has been performed for phase change, mixture and moving object. These analyses are depend on characteristics of that the scheme and the method are robust for steep change of density and useful for interface tracking. (author)

  9. Development of design technology on thermal-hydraulic performance in tight-lattice rod bundle. III - Numerical estimation on rod bowing effect based on X-ray CT data

    International Nuclear Information System (INIS)

    Misawa, Takeharu; Ohnuki, Akira; Katsuyama, Kozo; Nagamine, Tsuyoshi; Nakamura, Yasuo; Akimoto, Hajime; Mitsutake, Toru; Misawa, Susumu

    2007-01-01

    Design studies of the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) are being carried out at the Japan Atomic Energy Agency (JAEA) as one candidate for the future reactors. In actual core design, it is precondition to prevent fuel rods contact due to fuel rod bowing. However, the FLWR cores have nonconventional characteristics such as a hexagonal tight lattice arrangement and a high enrichment fuel loading. Therefore, as conservative evaluation, it is important to investigate influence of fuel rod bowing upon the boiling transition. In the JAEA, a 37-rod bundle experiments (base case test section (1.3mm gap width), gap width effect test section (1.0mm gap width), and rod bowing test section) were performed in order to investigate the thermal hydraulic characteristics in the tight lattice bundle. In this paper, the rod bowing effect test is paid attention. It is suspected that the actual fuel rod positions in the rod bowing test section may be different from the design-based positions. Even a slight displacement from the design-based position of fuel rod may occur variation of flow area, and give influence upon the thermal hydraulic characteristics in the rod bundle. Therefore, if the critical power in the rod bundle is evaluated by an analytical approach, the analysis based on more correct input can be performed by using actual fuel rod position data. In this study, the rod positions in the rod bowing test section were measured using the high energy X-ray computer tomography (Xray-CT). Based on the measured rod positions data, the subchannel analysis by the NASCA code was performed, in order to investigate applicability of the NASCA code to BT estimation of the rod bowing test section, and influence of displacement from design-based rod position upon BT estimation by the NASCA code. The predicted critical powers are agreement with those obtained by the experiment. The analysis based on the design-based rod positions is also performed, and the result is

  10. Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes

    International Nuclear Information System (INIS)

    Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.

    2003-01-01

    The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power

  11. Cross-cutting european thermal-hydraulics research for innovative nuclear systems

    International Nuclear Information System (INIS)

    Roelofs, F.; Class, A.; Cheng, X.; Meloni, P.; Van Tichelen, K.; Boudier, P.; Prasser, M.

    2010-01-01

    Thermal-hydraulics is recognized as a key scientific subject in the development of different innovative nuclear reactor systems. From the thermal-hydraulic point of view, different innovative reactors are mainly characterized by their coolants (gas, water, liquid metals and molten salt). This results in different micro- and macroscopic behavior of flow and heat transfer and requires specific models and advanced analysis tools. However, many common thermal-hydraulic issues are identified among various innovative nuclear systems. In Europe, such cross-cutting thermal-hydraulic issues are the subject of the 7. framework programme THINS (Thermal-Hydraulics of Innovative Nuclear Systems) project which runs from 2010 until 2014. This paper will describe the activities in this project which address the main identified thermal hydraulics issues for innovative nuclear systems. (authors)

  12. Thermal-hydraulics design comparisons for the tandem mirror hybrid reactor blanket

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Yang, Y.S.; Schultz, K.R.

    1980-09-01

    The Tandem Mirror Hybrid Reactor (TMHR) is a cylindrical reactor, and the fertile materials and tritium breeding fuel elements can be arranged with radial or axial orientation in the blanket module. Thermal-hydraulics performance comparisons were made between plate, axial rod and radial rod fuel geometrices. The three configurations result in different coolant/void fractions and different clad/structure fractions. The higher void fraction in the two rod designs means that these blankets will have to be thicker than the plate design blanket in order to achieve the same level of nuclear interactions. Their higher structural fractions will degrade the uranium breeding ratio and energy multiplication factor of the design. One difficulty in the thermal-hydraulics analysis of the plate design was caused by the varying energy multiplication of the blanket during the lifetime of the plate which forced the use of designs that operated in the transition flow regime at some point during life. To account for this, an approach was adopted from Gas Cooled Fast Reactor (GCFR) experience for the pressure drop calculation and the corresponding heat transfer coefficient that was used for the film drop thermal calculation. Because of the superior nuclear performance, the acceptable thermal-hydraulic characteristics and the mechanical design feasibility, the plate geometry concept was chosen for the reference gas-cooled TMHR blanket design

  13. 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.

  14. Thermal-hydraulic modeling of porous bed reactors

    International Nuclear Information System (INIS)

    Araj, K.J.; Nourbakhsh, H.P.

    1987-01-01

    Optimum design of nuclear reactor cores requires an iterative approach between the thermal-hydraulic, neutronic, and operational analysis. This paper will concentrate on the thermal-hydraulic behavior of a hydrogen-cooled small particle bed reactor (PBR). The PBR core modeled here consists of a hexagonal array of fuel elements embedded in a moderator matrix. The fuel elements are annular packed beds of fuel particles held between two porous cylindrical frits. These particles, 500 to 600 μm in diameter, have a uranium carbide core, which is coated by two layers of graphite and an outer coating of zirconium carbide. Coolant flows, radially inward, from the cold frit through the packed bed and hot frit and axially out the channel, formed by the hot frit to a common plenum. A fast running one-dimensional lumped-parameter steady-state code (FTHP) was developed to evaluate the effects of design changes in fuel assembly and power distribution. Another objective for the code was to investigate various methods of coolant control to minimize hot channel effects and maximize outlet temperatures

  15. Thermal hydraulic analysis of BWR containment venting system

    International Nuclear Information System (INIS)

    Baburajan, P.K.; Sharma, Prashant; Paul, U.K.; Gaikwad, Avinash

    2015-01-01

    Installation of additional containment filtered venting system (CFVS) is necessary to depressurize the containment to maintain its mechanical integrity due to over pressurization during severe accident condition. A typical venting system for BWR is modelled using RELAP5 and analysed to investigate the effect of various thermal hydraulic parameters on the operational parameters of the venting system. The venting system consists of piping from the containment to the scrubber tank and exit line from the scrubber tank. The scrubber tank is partially filled with water to enable the scrubbing action to remove the particulate radionuclides from the incoming containment air. The pipe line from the containment is connected to the venturi inlet and the throat of the venturi is open to the scrubber tank water inventory at designed submergence level. The exit of the venturi is open to scrubber tank water. Filters are used in the upper air space of the scrubber tank as mist separator before venting out the air into the atmosphere through the exit vent line. The effect of thermal hydraulic parameters such as inlet fluid temperature, inlet steam content and venturi submergence in the scrubber tank on the venting flow rate, exit steam content, scrubber tank inventory, overflow line and siphon breaker flow rate is analysed. Results show that inlet steam content and the venturi nozzle submergence influence the venting system parameters. (author)

  16. LWR containment thermal hydraulic codes benchmark demona B3 exercise

    International Nuclear Information System (INIS)

    Della Loggia, E.; Gauvain, J.

    1988-01-01

    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

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

    International Nuclear Information System (INIS)

    Staedtke, H.

    2001-01-01

    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)

  18. Scaling of Thermal-Hydraulic Phenomena and System Code Assessment

    International Nuclear Information System (INIS)

    Wolfert, K.

    2008-01-01

    In the last five decades large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Many separate effects tests and integral system tests were carried out to establish a data base for code development and code validation. In this context the question has to be answered, to what extent the results of down-scaled test facilities represent the thermal-hydraulic behaviour expected in a full-scale nuclear reactor under accidental conditions. Scaling principles, developed by many scientists and engineers, present a scientific technical basis and give a valuable orientation for the design of test facilities. However, it is impossible for a down-scaled facility to reproduce all physical phenomena in the correct temporal sequence and in the kind and strength of their occurrence. The designer needs to optimize a down-scaled facility for the processes of primary interest. This leads compulsorily to scaling distortions of other processes with less importance. Taking into account these weak points, a goal oriented code validation strategy is required, based on the analyses of separate effects tests and integral system tests as well as transients occurred in full-scale nuclear reactors. The CSNI validation matrices are an excellent basis for the fulfilling of this task. Separate effects tests in full scale play here an important role.

  19. Teaching Thermal Hydraulics & Numerical Methods: An Introductory Control Volume Primer

    Energy Technology Data Exchange (ETDEWEB)

    D. S. Lucas

    2004-10-01

    A graduate level course for Thermal Hydraulics (T/H) was taught through Idaho State University in the spring of 2004. A numerical approach was taken for the content of this course since the students were employed at the Idaho National Laboratory and had been users of T/H codes. The majority of the students had expressed an interest in learning about the Courant Limit, mass error, semi-implicit and implicit numerical integration schemes in the context of a computer code. Since no introductory text was found the author developed notes taught from his own research and courses taught for Westinghouse on the subject. The course started with a primer on control volume methods and the construction of a Homogeneous Equilibrium Model (HEM) (T/H) code. The primer was valuable for giving the students the basics behind such codes and their evolution to more complex codes for Thermal Hydraulics and Computational Fluid Dynamics (CFD). The course covered additional material including the Finite Element Method and non-equilibrium (T/H). The control volume primer and the construction of a three-equation (mass, momentum and energy) HEM code are the subject of this paper . The Fortran version of the code covered in this paper is elementary compared to its descendants. The steam tables used are less accurate than the available commercial version written in C Coupled to a Graphical User Interface (GUI). The Fortran version and input files can be downloaded at www.microfusionlab.com.

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

    International Nuclear Information System (INIS)

    Laffont, A.; Pentori, B.

    2003-01-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Laffont, A.; Pentori, B. [EDF R and D, EDF SEPTEN Electricity of France - Research and Development, Department SINETICS, 92 - Clamart (France)

    2003-07-01

    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)

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

    International Nuclear Information System (INIS)

    1991-01-01

    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.)

  3. Full scale mock-up tests for rod bundle thermal-hydraulics in Japan

    International Nuclear Information System (INIS)

    Sugawara, S.

    1995-01-01

    This poster describes tests aimed at development and validation of principal design methodology of rod bundle thermal-hydraulics correlations. The works are based on domestic data base using the full-scale mock-up test facilities. The scope of the tests comprises DNB heat flux, transient DNB heat flux, post DNB heat transfer, pressure drop and void distribution. The works have been performed under collaboration among electric facilities, NPP vendors, universities, governmental corporations. 1 tab., 14 figs

  4. The thermal-hydraulic for the new technologies: the micro-fluidics

    International Nuclear Information System (INIS)

    Crecy, F. de; Gruss, A.; Bricard, A.; Excoffon, J.

    2000-01-01

    The micro-fluidics can be defined as the fluid flow in little canals. This scale offers a great interest for the biotechnology type. In this paper, the authors present this fluids form and detail the researches performed at the Department of Physics and Thermal-hydraulics of the CEA, in the domain of the physical properties characterization and of the numerical two-phase direct simulation. (A.L.B.)

  5. Power transients of Ghana research reactor-1 using PARET/ANL thermal hydraulic code

    International Nuclear Information System (INIS)

    Ampomah-Amoaka, E.; Akaho, E.H.K.; Anim-Sampong, S.; Nyarko, B.J.B.

    2010-01-01

    PARET/ANL(Version 7.3 of 2007) thermal-hydraulic code was used to perform transient analysis of the Ghana Research Reactor-1.The reactivities inserted were 2.1mk and 4mk.The peak power of 5.81kW was obtained for 2.1 mk insertion whereas the peak power for 4mk insertion of reactivity was 92.32kW.These results compare closely with experiments and theoretical studies conducted previously.

  6. Loss-of-Fluid Test findings in pressurized water reactor core's thermal-hydraulic behavior

    International Nuclear Information System (INIS)

    Russell, M.

    1983-01-01

    This paper summarizes the pressurized water reactor (PWR) core's thermal-hydraulic behavior findings from experiments performed at the Loss-of-Fluid Test (LOFT) Facility at the Idaho National Engineering Laboratory. The potential impact of these findings on the safety and economics of PWR's generation of electricity is also discussed. Reviews of eight important findings in the core's physical behavior and in experimental methods are presented with supporting evidence

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

    International Nuclear Information System (INIS)

    Rodack, T.; Joffre, P.F.; Kapoor, R.K.

    2004-01-01

    ABB CE's improved System 80 TM 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 80 TM 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 80 TM fuel can be implemented with no net impact on Departure from Nucleate Boiling (DNB) margin in transition cores. (author)

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

    International Nuclear Information System (INIS)

    Vitkova, M.; Kalchev, B.; Stefanova, S.

    2006-01-01

    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

  9. Transitioning from interpretive to predictive in thermal hydraulic codes

    International Nuclear Information System (INIS)

    Mousseau, V.A.

    2004-01-01

    The current thermal hydraulic codes in use in the US, RELAP and TRAC, where originally written in the mid to late 1970's. At that time computers were slow, expensive, and had small memories. Because of these constraints, sacrifices had to be made, both in physics and numerical methods, which resulted in limitations on the accuracy of the solutions. Significant changes have occurred that induce very different requirements for the thermal hydraulic codes to be used for the future GEN-IV nuclear reactors. First, computers speed and memory grow at an exponential rate while the costs hold constant or decrease. Second, passive safety systems in modern designs stretch the length of relevant transients to many days. Finally, costs of experiments have grown very rapidly. Because of these new constraints, modern thermal hydraulic codes will be relied on for a significantly larger portion of bringing a nuclear reactor on line. Simulation codes will have to define in which part of state space experiments will be run. They will then have to be able to extend the small number of experiments to cover the large state space in which the reactors will operate. This data extrapolation mode will be referred to as 'predictive'. One of the keys to analyzing the accuracy of a simulation is to consider the entire domain being simulated. For example, in a reactor design where the containment is coupled to the reactor cooling system through radiative heat transfer, the accuracy of a transient includes the containment, the radiation heat transfer, the fluid flow in the cooling system, the thermal conduction in the solid, and the neutron transport in the reactor. All of this physics is coupled together in one nonlinear system through material properties, cross sections, heat transfer coefficients, and other mechanisms that exchange mass, momentum, and energy. Traditionally, these different physical domains, (containment, cooling system, nuclear fuel, etc.) have been solved in different

  10. An assessment of the CORCON-MOD3 code. Part 1: Thermal-hydraulic calculations

    International Nuclear Information System (INIS)

    Strizhov, V.; Kanukova, V.; Vinogradova, T.; Askenov, E.; Nikulshin, V.

    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

  11. Coupled neutronic-thermal-hydraulics analysis in a coolant subchannel of a PWR using CFD techniques

    Energy Technology Data Exchange (ETDEWEB)

    Ribeiro, Felipe P.; Su, Jian, E-mail: sujian@nuclear.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2017-07-01

    The high capacity of Computational Fluid Dynamics code to predict multi-dimensional thermal-hydraulics behaviour and the increased availability of capable computer systems are making that method a good tool to simulate phenomena of thermal-hydraulics nature in nuclear reactors. However, since there are no neutron kinetics models available in commercial CFD codes to the present day, the application of CFD in the nuclear reactor safety analyses is still limited. The present work proposes the implementation of the point kinetics model (PKM) in ANSYS - Fluent to predict the neutronic behaviour in a Westinghouse Sequoyah nuclear reactor, coupling with the phenomena of heat conduction in the rod and thermal-hydraulics in the cooling fluid, via the reactivity feedback. Firstly, a mesh convergence and turbulence model study was performed, using the Reynolds-Average Navier-Stokes method, with square arrayed rod bundle featuring pitch to diameter ratio of 1:32. Secondly, simulations using the k-! SST turbulence model were performed with an axial distribution of the power generation in the fuel to analyse the heat transfer through the gap and cladding, and its in fluence on the thermal-hydraulics behaviour of the cooling fluid. The wall shear stress distribution for the centre-line rods and the dimensionless velocity were evaluated to validate the model, as well as the in fluence of the mass flow rate variation on the friction factor. The coupled model enabled to perform a dynamic analysis of the nuclear reactor during events of insertion of reactivity and shutdown of primary coolant pumps. (author)

  12. Neutronics and thermal-hydraulics analysis of KUHFR

    Energy Technology Data Exchange (ETDEWEB)

    Woodruff, W L [Argonne National Laboratory, Argonne, IL (United States); Mishima, K [KURRI, Osaka (Japan)

    1983-08-01

    control rod worth with reduced enrichment has not yet determined, but only a small decrease in worth is expected. These BOL boron poisoned fuels are also used as the fresh fuel feed for the equilibrium fuel cycle studies contained in this report. The first three cases shown have matching cycle lengths in the equilibrium cycle, while the last case has a considerably longer cycle length. These results are similarly reflected in the 'Maximum Cycle Lengths' shown for unpoisoned BOL cores. Thus, the first three case can be considered comparable. The last case might be considered as an option for an extended cycle length design. The cycle length for this case is increased by about 21%. Obviously, by decreasing the uranium density in the fuel meat (to 2.7 g/cm{sup 3}), the cycle length for this design could be reduced to match that of the other cases. Thermal-hydraulic calculations have been carried out in order to study the safety aspects of the use of reduced enrichment uranium fuel for the KUHFR. The calculations were based on what is outlined in the Safety Analysis Report for the KUHFR and also the IAEA Guidebook for the RERTR program. Only a few combinations of hydraulic parameters have been tested because the reactor safety cannot be discussed without any nuclear physics considerations. For example, any variations in fuel coolant channels may change not only flow velocities but also power peaking factors which may affect the assessment of reactor safety. For this reason, the thermal-hydraulic calculations were carried out only for those specific cases on which neutronics analysis has been already performed. Low enriched uranium (LEU) cases are also included in this study as initial feasibility studies for potential conversion. The computer code PLTEMP has been developed to calculate the flow distribution in the core, fuel plate temperatures and DNB heat fluxes.

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

    International Nuclear Information System (INIS)

    Passalacqua, R.

    1994-01-01

    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)

  14. Thermal hydraulic analyses of LVR-15 research reactor with IRT-M fuel

    International Nuclear Information System (INIS)

    Macek, J.

    1997-01-01

    The LVR-15 pool-type research reactor has been in operation at the Nuclear Research Institute at Rez since 1955. Following a number of reconstructions and redesigning, the current reactor power is 15 MW. Thermal hydraulic analyses to demonstrate that the core heat will be safely removed during operation as well as in accident situations were performed based on methodology which had been specifically developed for the LVR-15 research reactor. This methodology was applied to stationary thermal hydraulic computations, as well as to transients, particularly with reactivity failure and loss of circulation pumps emergencies. The applied methodology and the core configuration as used in the Safety Report are described. The initial and boundary conditions are then considered and the summary of the calculated failures with regard to the defined safety limits is presented. The results of the core configuration analyses are also discussed with respect to meeting the safety limits and to the applicability of the methodology to this purpose

  15. A new approach to designing reduced scale thermal-hydraulic experiments

    International Nuclear Information System (INIS)

    Lapa, Celso M.F.; Sampaio, Paulo A.B. de; Pereira, Claudio M.N.A.

    2004-01-01

    Reduced scale experiments are often employed in engineering because they are much cheaper than real scale testing. Unfortunately, though, it is difficult to design a thermal-hydraulic circuit or equipment in reduced scale capable of reproducing, both accurately and simultaneously, all the physical phenomena that occur in real scale and operating conditions. This paper presents a methodology to designing thermal-hydraulic experiments in reduced scale based on setting up a constrained optimization problem that is solved using genetic algorithms (GAs). In order to demonstrate the application of the methodology proposed, we performed some investigations in the design of a heater aimed to simulate the transport of heat and momentum in the core of a pressurized water reactor (PWR) at 100% of nominal power and non-accident operating conditions. The results obtained show that the proposed methodology is a promising approach for designing reduced scale experiments

  16. Thermal-hydraulic design concept of the solid-target system of spallation neutron source

    International Nuclear Information System (INIS)

    Tanaka, F.; Hibiki, T.; Saito, Y.; Takeda, T.; Mishima, K.

    2001-01-01

    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)

  17. Condensation heat transfer coefficient with noncondensible gases for heat transfer in thermal hydraulic codes

    International Nuclear Information System (INIS)

    Banerjee, S.; Hassan, Y.A.

    1995-01-01

    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

  18. 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.

  19. 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.

  20. ATHENA [Advanced Thermal Hydraulic Energy Network Analyzer] solutions to developmental assessment problems

    International Nuclear Information System (INIS)

    Carlson, K.E.; Ransom, V.H.; Roth, P.A.

    1987-03-01

    The ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) code has been developed to perform transient simulation of the thermal hydraulic systems that may be found in fusion reactors, space reactors, and other advanced systems. As an assessment of current capability the code was applied to a number of physical problems, both conceptual and actual experiments. Results indicate that the numerical solution to the basic conservation equations is technically sound, and that generally good agreement can be obtained when modeling relevant hydrodynamic experiments. The assessment also demonstrates basic fusion system modeling capability and verifies compatibility of the code with both CDC and CRAY mainframes. Areas where improvements could be made include constitutive modeling, which describes the interfacial exchange term. 13 refs., 84 figs

  1. 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.

  2. Thermal-Hydraulic Integral Effect Test with the ATLS for Investigation on CEDM Penetration Nozzle Integrity

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoungho; Seokcho; Park, Hyunsik; Choi, Namhyun; Park, Yusun; Kim, Jongrok; Bae, Byounguhn; Kim, Yeonsik; Choi, Kiyong; Song, Chulhwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    In this study, thermal-hydraulic integral effect test with the ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation) was performed for simulating a failure of CEDM penetration nozzle. The main objectives of the present test were not only to provide physical insight into the system response during a failure of CEDM penetration nozzle but also to establish an integral effect test database for the validation of the safety analysis codes. Furthermore, present experimental data were utilized to resolve the safety issue raised by the PWSCC at the CEDM penetration nozzle of the YGN-3. Thermal-hydraulic integral effect test with the ATLAS was performed for simulating a failure of CEDM penetration nozzle. Failure of two penetration nozzles of the CEDM in the APR1400 was simulated. Initial and boundary conditions were determined with respect to the reference conditions of the APR1400. However, with an aim of corresponding to the YGN-3 situation, the safety injection water was supplied via CLI mode. Compared to the cold leg break SBLOCA, the consequences of the event were milder in terms of a loop seal clearance, break flow rate, collapsed water level, and PCT. This could be mainly attributed to the small break flow rate in case of the failure in the RPV upper head. Present experimental data were utilized to resolve the safety issue raised by the PWSCC at the CEDM penetration nozzle of the YGN-3.

  3. Thermal hydraulic feasibility analysis of the IBED PHTS for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Carloni, Dario, E-mail: dacarloni@gmail.com [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Pisa University, Via Diotisalvi 2, 56126 Pisa (Italy); Dell’Orco, Giovanni; Babulal, Gopalapillai; Somboli, Fabio; Serio, Luigi [ITER Organisation, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Paci, Sandro [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Pisa University, Via Diotisalvi 2, 56126 Pisa (Italy)

    2013-10-15

    One of the main challenges of the ITER fusion reactor is to effectively remove large amount of heat deposited to the surface of the plasma facing components. The tokamak cooling water system (TCWS) will accomplish the objective of removing about 1 GW of peak heat load from in-vessel components while maintaining pressures and temperatures of the coolant within acceptable and safe limits during different operational scenarios. A study of feasibility has been launched for the IBED PHTS (Integrated Blanket, Edge localized mode coils (ELMs) and Divertor Primary Heat Transfer System; it consists of five independent cooling trains (four operational and one in stand-by), one steam pressurizer, supply and return headers, ring manifolds and connections to the all in-vessel components (i.e. First Wall Blanket, Divertor, ELM, Diagnostics and other Ports clients). The dynamic behaviour of the IBED PHTS has been investigated by means of RELAP5{sup ®} code to simulate the response of the system during plasma pulse and baking operations. Due to the plasma heat deposition on the surfaces of the in-vessel components and subsequent increase in hot leg temperature, a large amount of water volume is transferred from the hot legs of the circuit to the surge-line of the pressurizer during each burn cycle. This causes rapid increase of pressure and temperature of the system and the following actions are proposed to counteract these variations: spray injection in the upper dome of the pressurizer from the Chemical and Volume Control System (CVCS) to reduce the pressure and active control of flow rates through heat exchangers and their bypass loops to regulate the heat transfer from the primary system to the environment via secondary and tertiary loops. This paper focuses on the prediction of the thermal hydraulic behaviour of the IBED PHTS during plasma pulses and baking scenarios, describing the various activity of the analysis, the geometrical assessment of the circuit and the modelling

  4. Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

    Energy Technology Data Exchange (ETDEWEB)

    O' Brien, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yoon, Su -Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States); Housley, Gregory K. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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

  5. Thermal hydraulic and neutronic interaction in the rotating bed reactor

    International Nuclear Information System (INIS)

    Lee, C.C.

    1986-01-01

    Power transient characteristics in a rotating fluidized bed reactor (RBR) are investigated theoretically. A propellant flow perturbation is assumed to occur in an initially equilibrium state of the core. Transfer functions representing quasi-one-dimensional mutual feedback between thermal hydraulics and neutronics are developed and analyzed in the frequency domain. Neutronic responses are determined by Fermi-age theory for slowing down of fast neutrons and diffusion theory for thermal neutron distribution. Neutron leakage through the exhaust nozzle is accounted for by applying diffuse view factors similar to those applied in radiative heat transfer. The bed expansion behavior is described by a kinematic wave equation derived from the continuity of the gas phase. The drift flux approach is used to determine the yield fractions in the equilibrium bed. Thermal responses of fuel are evaluated by dividing it into several volume-averaged zones to better account for the transient effects over single zone models. Sample calculations are undertaken for the various operation conditions and design parameters of the RBR based on 250 MW/sub t/, 1000 MW/sub t/, and 5000 MW/sub t/ power reactors. The results show that power transients are dependent on the parametric changes of optical thickness and view factors

  6. Development of thermal hydraulic analysis code for IHX of FBR

    International Nuclear Information System (INIS)

    Kumagai, Hiromichi; Naohara, Nobuyuki

    1991-01-01

    In order to obtain flow resistance correlations for thermal-hydrauric analysis code concerned with an intermediate heat exchanger (IHX) of FBR, the hydraulic experiment by air was carried out through a bundle of tubes arranged in an in-line and staggard fashion. The main results are summarized as follows. (1) On pressure loss per unit length of a tube bundle, which is densely a regular triangle arrangement, the in-line fashion is almost the same as the staggard one. (2) In case of 30deg sector model for IHX tube bundle, pressure loss is 1/3 in comparison with the in-line or staggard arrangement. (3) By this experimental data, flow resistance correlations for thermalhydrauric analysis code are obtained. (author)

  7. Evolution of thermal-hydraulics testing in EBR-II

    International Nuclear Information System (INIS)

    Golden, G.H.; Planchon, H.P.; Sackett, J.I.; Singer, R.M.

    1987-01-01

    A thermal-hydraulics testing and modeling program has been underway at the Experimental Breeder Reactor-II (EBR-II) for 12 years. This work culminated in two tests of historical importance to commercial nuclear power, a loss of flow without scram and a loss of heat sink wihout scram, both from 100% initial power. These tests showed that natural processes will shut EBR-II down and maintain cooling without automatic control rod action or operator intervention. Supporting analyses indicate that these results are characteristic of a range of sizes of liquid metal cooled reactors (LMRs), if these reactors use metal driver fuel. This type of fuel is being developed as part of the Integral Fast Reactor Program at Argonne National Laboratory. Work is now underway at EBR-II to exploit the inherent safety of metal-fueled LMRs with regard to development of improved plant control strategies. (orig.)

  8. Thermal hydraulic stability in a pressure tube nuclear reactor

    International Nuclear Information System (INIS)

    Villani, A.; Ravetta, R.; Mansani, L.

    1986-01-01

    The CIRENE plant which will undergo preoperational tests in the near future is equipped with a 40 MW(e) Heavy Water moderated Boiling Light Water cooled Reactor (HWBLWR); at the start-up and up to about 30 % of nominal power, the necessary low coolant density is obtained injecting into the core a mixture of liquid and steam. To verify the thermal-hydraulic stability of the plant in this situation, tests have been carried out in a facility simulating two full scale power channels; the system stability has been confirmed in the reference conditions, and is not reduced by even a significant reduction of the liquid flowrate, where a decrease in liquid temperature has some negative effect and steam flowrate has a small influence. (author)

  9. GNPS 18-months fuel cycles core thermal hydraulic design

    International Nuclear Information System (INIS)

    Liu Changwen; Zhou Zhou

    2002-01-01

    GNPS begins to implement the 18-month fuel cycles from the initial annual reload at cycle 9, thus the initial core thermal hydraulic design is not valid any more. The new critical heat flux (CHF) correlation, FC, which is developed by Framatome, is used in the design, and the generalized statistical methodology (GSM) instead of the initial deterministic methodology is used to determine the DNBR design limit. As the AFA 2G and AFA 3G are mixed loaded in the transition cycle, it will result that the minimum DNBR in the mixed core is less than that of AFA 3G homogenous core, the envelop mixed core DNBR penalty is given. Consequently the core physical limit for mixed core and equilibrium cycles, and the new over temperature ΔT overpower ΔT are determined

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

    International Nuclear Information System (INIS)

    Copp, H.W.; Shashidhara, N.S.

    1981-01-01

    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

  11. Report on the thermal-hydraulics computational component

    International Nuclear Information System (INIS)

    Laughton, T.; Jones, B.G.

    1996-01-01

    The nodal methods computer code utilizing hexagonal geometry, which is being developed as part of this DOE contract, is called THMZ. The computational objective of the code is to calculate the steady-state thermal-hydraulic conditions in a hexagonal geometry reactor core given the appropriate initial conditions and the axial neutron flux profile. The latter is given by a companion nodal neutronics code which was developed in an earlier part of the contact. The joining of these two codes to provide a coupled analysis tool for hexagonal lattice cores is the ultimate objective of the contract and its follow-on work. The remaining part of this report presents the current status of the development and the results which have been obtained to date. These will appear in the MS thesis of Mr. Terrill Laughton in the Department of Nuclear Engineering which is currently in preparation

  12. Validation of the TEXSAN thermal-hydraulic analysis program

    International Nuclear Information System (INIS)

    Burns, S.P.; Klein, D.E.

    1992-01-01

    The TEXSAN thermal-hydraulic analysis program has been developed by the University of Texas at Austin (UT) to simulate buoyancy driven fluid flow and heat transfer in spent fuel and high level nuclear waste (HLW) shipping applications. As part of the TEXSAN software quality assurance program, the software has been subjected to a series of test cases intended to validate its capabilities. The validation tests include many physical phenomena which arise in spent fuel and HLW shipping applications. This paper describes some of the principal results of the TEXSAN validation tests and compares them to solutions available in the open literature. The TEXSAN validation effort has shown that the TEXSAN program is stable and consistent under a range of operating conditions and provides accuracy comparable with other heat transfer programs and evaluation techniques. The modeling capabilities and the interactive user interface employed by the TEXSAN program should make it a useful tool in HLW transportation analysis

  13. A review of the current thermal-hydraulic modeling of the Jules Horowitz Reactor: A loss of flow accident analysis

    International Nuclear Information System (INIS)

    Pegonen, R.; Bourdon, S.; Gonnier, C.; Anglart, H.

    2014-01-01

    Highlights: • CEA methodology for thermal-hydraulic calculations in the JHR reactor is described. • Thermal-hydraulics of the JHR is analyzed during LOFA using CATHARE and FLICA4. • Safety criteria, important modeling parameters and correlations are presented. • Possible improvements of the current methodology are discussed and proposed. - Abstract: The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support existing and future nuclear reactor designs. The reactor is under construction at CEA Cadarache research center in France and is expected to start operation at the end of this decade. R and D and analytical works have already been performed to set-up the methodology for thermal-hydraulic calculations of the reactor. This paper presents the off-line coupled thermal-hydraulic modeling of the reactor using the CATHARE system code and the FLICA4 core analysis code. The main objective of the present work is to analyze the thermal-hydraulic calculations of the reactor during the loss of flow accident using CEA methodology. Possible improvements of the current methodology are shortly discussed and suggested

  14. Status and subjects of thermal-hydraulic analysis for next-generation LWRs with passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    The present status and subjects on thermal-hydraulic analysis for next-generation light water reactors (LWRs) with passive safety systems were summarized based on survey results and discussion by subcommittee on improvement of reactor thermal-hydraulic analysis codes under nuclear code committee in Japan Atomic Energy Research Institute. This survey was performed to promote the research of improvement of reactor thermal-hydraulic analysis codes in future. In the first part of this report, the status and subjects on system analysis and those on evaluation of passive safety system performance are summarized for various types of reactor proposed before. In the second part, the status and subjects on multidimensional two-phase flow analysis are reviewed, since the multidimensional analysis was recognized as one of most important subjects through the investigation in the first part. Besides, databases for bubbly flow and annular dispersed flow were explored, those are needed to assess and verify each multidimensional analytical method. The contents in this report are the forefront of thermal-hydraulic analysis for LWRs and those include current findings for the development of multidimensional two-phase flow analytical method. Thus, we expect that the contents can offer various useful information against the improvement of reactor thermal-hydraulic analysis codes in future. (author)

  15. R and D on thermal hydraulics of core and core-bottom structure

    International Nuclear Information System (INIS)

    Inagaki, Yoshiyuki; Hino, Ryutaro; Kunitomi, Kazuhiko; Takase, Kazuyuki; Ioka, Ikuo; Maruyama, So

    2004-01-01

    Thermal hydraulic tests on the core and core-bottom structure of the high-temperature engineering test reactor (HTTR) were carried out with the helium engineering demonstration loop (HENDEL) under simulated reactor operating conditions. The HENDEL was composed of helium gas circulation loops, mother sections (M 1 and M 2 ) and adaptor section (A), and two test sections, i.e. the fuel stack test section (T 1 ) and in-core structure test section (T 2 ). In the T 1 test section simulating a fuel stack of the core, thermal and hydraulic performances of helium gas flowing through a fuel block were investigated for thermal design of the HTTR core. In the T 2 test section simulating the core-bottom structure, demonstration tests were performed to verify the structural integrity of graphite and metal components, seal performance against helium gas leakage among the graphite permanent blocks and thermal mixing performance of helium gas. The above test results in the T 1 and T 2 test sections were applied to the detailed design and licensing works of the HTTR and the HENDEL-loop was dismantled in 1999

  16. COOLOD, Steady-State Thermal Hydraulics of Research Reactors

    International Nuclear Information System (INIS)

    Kaminaga, Masanori

    1997-01-01

    1 - Description of program or function: The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is a revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode. A 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is a subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. 2 - Method of solution: The 'Heat Transfer Package' is a subprogram for calculating heat transfer coefficients, ONB temperature, heat flux at onset of flow instability and DNB heat flux. The 'Heat transfer package' was especially developed for research reactors which are operated under low pressure and low temperature conditions using plate-type fuel, just like the JRR-3M. Heat transfer correlations adopted in the 'Heat Transfer Package' were obtained or estimated based on the heat transfer experiments in which thermal-hydraulic features of the upgraded JRR-3 core were properly reflected. The 'Heat Transfer Package' is applicable to upward and downward flow

  17. ATLAS program for advanced thermal-hydraulic safety research

    International Nuclear Information System (INIS)

    Song, Chul-Hwa; Choi, Ki-Yong; Kang, Kyoung-Ho

    2015-01-01

    Highlights: • Major achievements of the ATLAS program are highlighted in conjunction with both developing advanced light water reactor technologies and enhancing the nuclear safety. • The ATLAS data was shown to be useful for the development and licensing of new reactors and safety analysis codes, and also for nuclear safety enhancement through domestic and international cooperative programs. • A future plan for the ATLAS testing is introduced, covering recently emerging safety issues and some generic thermal-hydraulic concerns. - Abstract: This paper highlights the major achievements of the ATLAS program, which is an integral effect test program for both developing advanced light water reactor technologies and contributing to enhancing nuclear safety. The ATLAS program is closely related with the development of the APR1400 and APR"+ reactors, and the SPACE code, which is a best-estimate system-scale code for a safety analysis of nuclear reactors. The multiple roles of ATLAS testing are emphasized in very close conjunction with the development, licensing, and commercial deployment of these reactors and their safety analysis codes. The role of ATLAS for nuclear safety enhancement is also introduced by taking some examples of its contributions to voluntarily lead to multi-body cooperative programs such as domestic and international standard problems. Finally, a future plan for the utilization of ATLAS testing is introduced, which aims at tackling recently emerging safety issues such as a prolonged station blackout accident and medium-size break LOCA, and some generic thermal-hydraulic concerns as to how to figure out multi-dimensional phenomena and the scaling issue.

  18. 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.

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

    International Nuclear Information System (INIS)

    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

  20. Investigation of coupling scheme for neutronic and thermal-hydraulic codes

    International Nuclear Information System (INIS)

    Wang Guoli; Yu Jianfeng; Pen Muzhang; Zhang Yuman.

    1988-01-01

    Recently, a number of coupled neutronics/thermal-hydraulics codes have been used in reaction design and safty analysis, which have been obtained by coupling previous neutronic and thermal-hydraulic codes. The different coupling schemes affect computer time and accuracy of calculation results. Numberical experiments of several different coupling schemes and some heuristic results are described

  1. Thermal-hydraulic simulation and analysis of Research Reactor Cooling Systems

    International Nuclear Information System (INIS)

    EL Khatib, H.H.A.

    2013-01-01

    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.

  2. Investigation and Development of the Thermal Preparation System of the Trailbuilder Machinery Hydraulic Actuator

    Science.gov (United States)

    Konev, V.; Polovnikov, E.; Krut, O.; Merdanov, Sh; Zakirzakov, G.

    2017-07-01

    It’s determined that the main part of trailbuilders operated in the North is the technology equipped by the hydraulic actuator. Further development of the northern territories will demand using of various means and ways machinery thermal preparation, and also the machinery of the northern fulfillment. On this basis problems in equipment operation are defined. One of the main is efficiency supplying of a hydraulic actuator. On the basis of the operating conditions’ analysis of trailbuilder hydraulic actuator operation it is determined, that under low negative temperatures the means of thermal preparation are necessary. The existing systems warm up only a hydraulic tank or warming up of the hydro equipment before the machinery operation is carried out under loading with intensive wears. Thus, with the purpose to raise the efficiency of thermal hydraulic actuator, operated far from stationary bases autonomous, energy saving, not expensive in creation and operation systems are necessary. In accordance with the analysis of means and ways of the thermal preparation of the hydraulic actuator and the thermal balance calculations of the (internal) combustion engine the system of the hydraulic actuator heating is offered and is being investigated. It contains a local hydraulic actuator warming up and the system of internal combustion engine heat utilization. Within research operation conditions of the local hydraulic actuator heating are viewed and determined, taking into account constructive changes to the local hydraulic actuator heating. Mathematical modelling of the heat technical process in the modernized hydraulic actuator is considered. As a result temperature changes of the heat-transfer and the hydraulic cylinder in time are determined. To check the theoretical researches and to define dependences on hydraulic actuator warming up, the experimental installation is made. It contains the measuring equipment, a small tank with the heat exchanger of the burnt gases

  3. 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.

  4. Neutronic and thermal-hydraulic studies of aqueous homogeneous reactor for medical isotopes production

    International Nuclear Information System (INIS)

    Perez, Daniel Milian; Lorenzo, Daniel E. Milian; Lira, Carlos A. Brayner de Oliveira; Garcia, Lorena P. Rodríguez; Universidade Federal de Pernambuco

    2017-01-01

    The use of Aqueous Homogenous Reactors (AHR) is one of the most promissory alternatives to produce medical isotopes, mainly "9"9Mo. Compare to multipurpose research reactors, an AHR dedicated for "9"9Mo production has advantages because of their low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing, and purification characteristics. This article presents the current state of research in our working group on this topic. Are presented and discussed the group validation efforts with benchmarking exercises that include neutronic and thermal-hydraulic results of two solution reactors, the SUPO and ARGUS reactors. Neutronic and thermal-hydraulic results of 75 kWth AHR based on the ARGUS reactor LEU configuration are presented. The neutronic studies included the determination of parameters such as reflector thickness, critical height, medical isotopes production and others. Thermal-hydraulics studies were focused on demonstrating that sufficient cooling capacity exists to prevent fuel overheating. In addition, the effects of some calculation parameters on the computational modeling of temperature, velocity and gas volume fraction during steady-state operation of an AHR are discussed. The neutronic and thermal-hydraulics studies have been performed with the MCNPX version 2.6e computational code and the version 14 of ANSYS CFX respectively. Our group studies and the results obtained contribute to demonstrate the feasibility of using AHR for the production of medical isotopes, however additional studies are still necessary to confirm these results and contribute to development and demonstration of their technical, safety, and economic viability. (author)

  5. Neutronic and thermal-hydraulic studies of aqueous homogeneous reactor for medical isotopes production

    Energy Technology Data Exchange (ETDEWEB)

    Perez, Daniel Milian; Lorenzo, Daniel E. Milian; Lira, Carlos A. Brayner de Oliveira; Garcia, Lorena P. Rodríguez, E-mail: milianperez89@gmail.com, E-mail: dmilian@instec.cu, E-mail: lorenapilar1109@gmail.com, E-mail: cabol@ufpe.br [Higher Institute of Technologies and Applied Sciences (InSTEC), Havana (Cuba); Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear

    2017-11-01

    The use of Aqueous Homogenous Reactors (AHR) is one of the most promissory alternatives to produce medical isotopes, mainly {sup 99}Mo. Compare to multipurpose research reactors, an AHR dedicated for {sup 99}Mo production has advantages because of their low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing, and purification characteristics. This article presents the current state of research in our working group on this topic. Are presented and discussed the group validation efforts with benchmarking exercises that include neutronic and thermal-hydraulic results of two solution reactors, the SUPO and ARGUS reactors. Neutronic and thermal-hydraulic results of 75 kWth AHR based on the ARGUS reactor LEU configuration are presented. The neutronic studies included the determination of parameters such as reflector thickness, critical height, medical isotopes production and others. Thermal-hydraulics studies were focused on demonstrating that sufficient cooling capacity exists to prevent fuel overheating. In addition, the effects of some calculation parameters on the computational modeling of temperature, velocity and gas volume fraction during steady-state operation of an AHR are discussed. The neutronic and thermal-hydraulics studies have been performed with the MCNPX version 2.6e computational code and the version 14 of ANSYS CFX respectively. Our group studies and the results obtained contribute to demonstrate the feasibility of using AHR for the production of medical isotopes, however additional studies are still necessary to confirm these results and contribute to development and demonstration of their technical, safety, and economic viability. (author)

  6. 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)

    1999-12-31

    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)

  7. TRSM-a thermal-hydraulic real-time simulation model for PWR

    International Nuclear Information System (INIS)

    Zhou Weichang

    1997-01-01

    TRSM (a Thermal-hydraulic Real-time Simulation Model) has been developed for PWR real-time simulation and best-estimate prediction of normal operating and abnormal accident conditions. It is a non-equilibrium two phase flow thermal-hydraulic model based on five basic conservation equations. A drift flux model is used to account for the unequal velocities of liquid and gaseous mixture, with or without the presence of the noncondensibles. Critical flow models are applied for break flow and valve flow calculations. A 5-regime two phase heat convection model is applied for clad-to-coolant as well as fluid-to-tubing heat transfer. A rigorous reactor coolant pump model is used to calculate the pressure drop and rise for the suction and discharge ends with complete pump characteristics curves included. The TRSM model has been adapted in the full-scale training simulator of Qinshan Nuclear Power Plant 300 MW unit to simulate the thermal-hydraulic performance of the NSSS. The simulation results of a cold leg LOCA and a steam generator tube rupture (SGTR) accident are presented

  8. Evaluation of operational safety at Babcock and Wilcox Plants: Volume 2, Thermal-hydraulic results

    International Nuclear Information System (INIS)

    Wheatley, P.D.; Davis, C.B.; Callow, R.A.; Fletcher, C.D.; Dobbe, C.A.; Beelman, R.J.

    1987-11-01

    The Nuclear Regulatory Commission has initiated a research program to develop a methodology to assess the operational performance of Babcock and Wilcox plants and to apply this methodology on a trial basis. The methodology developed for analyzing Babcock and Wilcox plants integrated methods used in both thermal-hydraulics and human factors and compared results with information used in the assessment of risk. The integrated methodology involved an evaluation of a selected plant for each pressurized water reactor vendor during a limited number of transients. A plant was selected to represent each vendor, and three transients were identified for analysis. The plants were Oconee Unit 1 for Babcock and Wilcox, H.B. Robinson Unit 2 for Westinghouse, and Calvert Cliffs Unit 1 for Combustion Engineering. The three transients were a complete loss of all feedwater, a small-break loss-of-coolant accident, and a steam-generator overfill with auxiliary feedwater. Included in the integrated methodology was an assessment of the thermal-hydraulic behavior, including event timing, of the plants during the three transients. Thermal-hydraulic results are presented in this volume (Volume 2) of the report. 26 refs., 30 figs., 7 tabs

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

    International Nuclear Information System (INIS)

    Macek, J.

    2001-01-01

    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. 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)

  11. Hydraulic and thermal conduction phenomena in soils at the particle-scale: Towards realistic FEM simulations

    International Nuclear Information System (INIS)

    Narsilio, G A; Yun, T S; Kress, J; Evans, T M

    2010-01-01

    This paper summarizes a method to characterize conduction properties in soils at the particle-scale. The method set the bases for an alternative way to estimate conduction parameters such as thermal conductivity and hydraulic conductivity, with the potential application to hard-to-obtain samples, where traditional experimental testing on large enough specimens becomes much more expensive. The technique is exemplified using 3D synthetic grain packings generated with discrete element methods, from which 3D granular images are constructed. Images are then imported into the finite element analyses to solve the corresponding governing partial differential equations of hydraulic and thermal conduction. High performance computing is implemented to meet the demanding 3D numerical calculations of the complex geometrical domains. The effects of void ratio and inter-particle contacts in hydraulic and thermal conduction are explored. Laboratory measurements support the numerically obtained results and validate the viability of the new methods used herein. The integration of imaging with rigorous numerical simulations at the pore-scale also enables fundamental observation of particle-scale mechanisms of macro-scale manifestation.

  12. Measurement and flow visualization research of thermal hydraulic characteristics for the SFR reactor Vessel

    International Nuclear Information System (INIS)

    Cha, J. E.; Kim, S. O.; Choi, H. L.; Kim, H. B.; Kim, H. W.; Lee, S. H.

    2012-01-01

    In this report, the thermal hydraulic and flow visualization experiment was described for the KALIMER-600 water-scaled model. In order to investigate a thermal hydraulic characteristics for the SFR KALIMER-600, which has been conceptually designed in the KAERI, a water-scaled 1/10 reactor vessel model was designed and prepared through the scaling analysis during three-years research. In this research, SFR Photos system, which has inherently very complicated the internal structures, was fabricated with a transparent vessel. It was shown that a serious of thermal hydraulic test was conducted within a short period if modeled with water than sodium. Natural circulation test was successfully performed with the modeled heater assembly and heat exchanger system coupled with cooling system. The water-scaled RSV experimental facility made in this research could be used to study the USA development for the future SFR system and utilized to analyze the flow characteristics before changing a main internal part of Photos system. It could also be used to test a pool-inspection study and a sensor selection study before large scale sodium experiment. The PCV system prepared in this research could be utilized to test other TSH experiment and temperature field measurement

  13. Application of MELCOR Code to a French PWR 900 MWe Severe Accident Sequence and Evaluation of Models Performance Focusing on In-Vessel Thermal Hydraulic Results

    International Nuclear Information System (INIS)

    De Rosa, Felice

    2006-01-01

    In the ambit of the Severe Accident Network of Excellence Project (SARNET), funded by the European Union, 6. FISA (Fission Safety) Programme, one of the main tasks is the development and validation of the European Accident Source Term Evaluation Code (ASTEC Code). One of the reference codes used to compare ASTEC results, coming from experimental and Reactor Plant applications, is MELCOR. ENEA is a SARNET member and also an ASTEC and MELCOR user. During the first 18 months of this project, we performed a series of MELCOR and ASTEC calculations referring to a French PWR 900 MWe and to the accident sequence of 'Loss of Steam Generator (SG) Feedwater' (known as H2 sequence in the French classification). H2 is an accident sequence substantially equivalent to a Station Blackout scenario, like a TMLB accident, with the only difference that in H2 sequence the scram is forced to occur with a delay of 28 seconds. The main events during the accident sequence are a loss of normal and auxiliary SG feedwater (0 s), followed by a scram when the water level in SG is equal or less than 0.7 m (after 28 seconds). There is also a main coolant pumps trip when ΔTsat < 10 deg. C, a total opening of the three relief valves when Tric (core maximal outlet temperature) is above 603 K (330 deg. C) and accumulators isolation when primary pressure goes below 1.5 MPa (15 bar). Among many other points, it is worth noting that this was the first time that a MELCOR 1.8.5 input deck was available for a French PWR 900. The main ENEA effort in this period was devoted to prepare the MELCOR input deck using the code version v.1.8.5 (build QZ Oct 2000 with the latest patch 185003 Oct 2001). The input deck, completely new, was prepared taking into account structure, data and same conditions as those found inside ASTEC input decks. The main goal of the work presented in this paper is to put in evidence where and when MELCOR provides good enough results and why, in some cases mainly referring to its

  14. Development of realistic thermal hydraulic system analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Chung, B. D; Kim, K. D. [and others

    2002-05-01

    The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others.

  15. Development of realistic thermal hydraulic system analysis code

    International Nuclear Information System (INIS)

    Lee, Won Jae; Chung, B. D; Kim, K. D.

    2002-05-01

    The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others

  16. Thermal Hydraulic Analysis of RPV Support Cooling System for HTGR

    International Nuclear Information System (INIS)

    Min Qi; Wu Xinxin; Li Xiaowei; Zhang Li; He Shuyan

    2014-01-01

    Passive safety is now of great interest for future generation reactors because of its reduction of human interaction and avoidance of failures of active components. reactor pressure vessel (RPV) support cooling system (SCS) for high temperature gas-cooled reactor (HTGR) is a passive safety system and is used to cool the concrete seats for the four RPV supports at its bottom. The SCS should have enough cooling capacity to ensure the temperature of the concrete seats for the supports not exceeding the limit temperature. The SCS system is composed of a natural circulation water loop and an air cooling tower. In the water loop, there is a heat exchanger embedded in the concrete seat, heat is transferred by thermal conduction and convection to the cooling water. Then the water is cooled by the air cooler mounted in the air cooling tower. The driving forces for water and air are offered by the density differences caused by the temperature differences. In this paper, the thermal hydraulic analysis for this system was presented. Methods for decoupling the natural circulation and heat transfer between the water loop and air flow were introduced. The operating parameters for different working conditions and environment temperatures were calculated. (author)

  17. Thermal-hydraulic analysis on the whole module of water cooled ceramic breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Kecheng; Ma, Xuebin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Cheng, Xiaoman [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Lin, Shuang [University of Science and Technology of China, Hefei, Anhui, 230027 (China); Huang, Kai [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China)

    2016-11-15

    Highlights: • The 3D thermal hydraulic analysis on the whole module of WCCB is performed by CFD method. • Temperature field and mass flow distribution have been obtained. • The design of WCCB is reasonable from the perspective of thermal-hydraulics. • The scheme for further optimization has been proposed. - Abstract: The Water Cooled Ceramic Breeder blanket (WCCB) is being researched for Chinese Fusion Engineering Test Reactor (CFETR). The thermal-hydraulic analysis is essential because the blanket should remove the high heat flux from the plasma and the volumetric heat generated by neutrons. In this paper, the detailed three dimensional (3D) thermal hydraulic analysis on the whole module of WCCB blanket has been performed by Computational Fluid Dynamics (CFD) method, which is capable of solving conjugate heat transfer between solid structure and fluid. The main results, including temperature field, distribution of mass flow rate and coolant pressure drop, have been calculated simultaneously. These provides beneficial guidance data for the further structural optimization and for the design arrangement of primary and secondary circuit. Under the total heat source of 1.23 MW, the coolant mass flow rate of 5.457 kg/s is required to make coolant water corresponding to the Pressurized Water Reactor (PWR) condition (15.5 MPa, 285 °C–325 °C), generating the total coolant pressure drop (△P) of 0.467 MPa. The results show that the present structural design can make all the materials effectively cooled to the allowable temperature range, except for a few small modifications on the both sides of FW. The main components, including the first wall (FW), cooling plates (CPs), side wall (SWs)&stiffening plates (SPs) and the manifold(1–4), dominate 4.7%/41.7%/13%/40.6% of the total pressure drop, respectively. Additionally, the mass flow rate of each channel has been obtained, showing the peak relative deviation of 3.4% and 2% from the average for the paratactic

  18. 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

  19. Comparative Studies of Core Thermal Hydraulic Design Methods for the Prototype Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Choi, Sun Rock; Lim, Jae Yong; Kim, Sang Ji

    2013-01-01

    In this work, various core thermal-hydraulic design methods, which have arisen during the development of a prototype SFR, are compared to establish a proper design procedure. Comparative studies have been performed to determine the appropriate design method for the prototype SFR. The results show that the minimization method show a lower cladding midwall temperature than the fixed outlet temperature methods and superior thermal safety margin with the same coolant flow. The Korea Atomic energy Research Institute (KAERI) has performed a conceptual SFR design with the final goal of constructing a prototype plant by 2028. The main objective of the SFR prototype plant is to verify the TRU metal fuel performance, reactor operation, and transmutation ability of high-level wastes. The core thermal-hydraulic design is used to ensure the safe fuel performance during the whole plant operation. Compared to the critical heat flux in typical light water reactors, nuclear fuel damages in SFR subassemblies are arisen from a creep induced failure. The creep limit is evaluated based on both the maximum cladding temperature and the uncertainties of the design parameters. Therefore, the core thermalhydraulic design method, which eventually determines the cladding temperature, is highly important to assure a safe and reliable operation of the reactor systems

  20. Analysis of the Phebus FPT0 containment thermal hydraulics with the Jericho and Trio-VF codes

    International Nuclear Information System (INIS)

    Layly, V.D.; Spitz, P.; Mailliat, A.

    1994-01-01

    This paper presents the analysis of the thermal hydraulic behavior of the containment, during the Phebus FPT0 test performed on December 2, 1993, with the Jericho code which deals with the thermal hydraulics of containment in the severe accident field. This code is part of Escadre which is the French system of codes in charge of predicting PWR severe accidents. After summarizing the relevant Jericho code characteristics and the preliminary assessment work for the Phebus conditions, we briefly describe the REPF 502 test facility and report the thermal hydraulic FPT0 experimental protocol. Then, the experiment / Jericho calculation comparisons are analysed. Because the Jericho code assumes a well-mixed atmosphere, some additional 3-D calculations have been carried out in order to get further insight on the convection flow patterns and qualify the well-mixed atmosphere assumption in the Phebus containment. (author). 9 refs., 12 figs

  1. Development of RETRAN-03/MOV code for thermal-hydraulic analysis of nuclear reactor under moving conditions

    International Nuclear Information System (INIS)

    Kim, Hak Jae; Park, Goon Cherl

    1996-01-01

    Nuclear ship reactors have several; features different from land-based PWR's. Especially, effects of ship motions on reactor thermal-hydraulics and good load following capability for abrupt load changes are essential characteristics of nuclear ship reactors. This study modified the RETRAN-03 to analyze the thermal-hydraulic transients under three-dimensional ship motions, named RETRAN-03/MOV in order to apply to future marine reactors. First Japanese nuclear ship MUTSU reactor have been analyzed under various ship motions to verify this code. Calculations have been performed under rolling,heaving and stationary inclination conditions during normal operation. Also, the natural circulation has been analyzed, which can provide the decay heat removed to ensure the passive safety of marine reactors. As results, typical thermal-hydraulic characteristics of marine reactors such as flow rate oscillations and S/G water level oscillations have been successfully simulated at various conditions. 7 refs., 11 figs. (author)

  2. Oyster Creek fuel thermal margin during core thermal-hydraulic oscillations

    International Nuclear Information System (INIS)

    Dougher, J.D.

    1990-01-01

    The Oyster Creek nuclear facility, a boiling water reactor (BWR)-2 plant type, has never experienced core thermal-hydraulic instability. Power oscillations, however, have been observed in other BWR cores both domestically and internationally. Two modes of oscillations have been observed, core wide and regional half-core. During core wide oscillations, the neutron flux in the core oscillates in the radial fundamental mode. During regional half-core oscillations, higher order harmonics in the radial plane result in out-of-phase oscillations with the neutron flux in one half of the core oscillating 180 deg out-of-phase with the neutron flux in the other half of the core. General Design Criteria 12 requires either prevention or detection and suppression of power oscillations which could result in violations of fuel design limits. Analyses performed by General Electric have demonstrated that for large-magnitude oscillations the potential exists for violation of the safety limit minimum critical power ratio (MCPR). However, for plants with a flow-biased neutron flux scram automatic mitigation of oscillations may be provided at an oscillation magnitude below that at which the safety limit is challenged. Plant-specific analysis for Oyster Creek demonstrates that the existing average power range monitor (APRM) system will sense and suppress power oscillations prior to violation of any safety limits

  3. Coupled neutronic and thermal-hydraulic code benchmark activities at the International Nuclear Safety Center

    International Nuclear Information System (INIS)

    Podlazov, L. N.

    1998-01-01

    Two realistic benchmark problems are defined and used to assess the performance of coupled thermal-hydraulic and neutronic codes used in simulating dynamic processes in VVER-1000 and RBMK reactor systems. One of the problems simulates a design basis accident involving the ejection of three control and protection system rods from a VVER-1000 reactor. The other is based on a postulated rod withdrawal from an operating RBMK reactor. Preliminary results calculated by various codes are compared. While these results show significant differences, the intercomparisons performed so far provide a basis for further evaluation of code limitations and modeling assumptions

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

    International Nuclear Information System (INIS)

    Krotiuk, W.J.

    1990-01-01

    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

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

    International Nuclear Information System (INIS)

    D'Auria, Francesco; Bousbia Salah, Anis; Galassi, G.M.; Vedovi, Juswald; Van Goethem, Georges; Hadek, Jan; Macek, Jiri; Rindelhardt, Udo; Rohde, Ulrich; Ahnert Iglesias, Carol; Aragones Beltran, Jose Maria; Reventos, Francesc; Cuadra, Arantxa; Gago, Jose Luis; Verdu, Gumersindo; Miro, Rafael; Ginestar, Damian; Sanchez, Ana Maria; Sjoberg, Anders; Yitbarek, M.; Sandervag, Oddbjoern; Garis, Ninos; Frid, Wiktor; Panayotov, Dobromir; Ivanov, Kostadin; Uddin, Rizwan; Sartori, Enrico

    2004-01-01

    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

  6. The Phebus FP thermal-hydraulic analysis with Melcor

    International Nuclear Information System (INIS)

    Akgane, Kikuo; Kiso, Yoshihiro; Fukahori, Takanori; Yoshino, Mamoru

    1995-01-01

    The severe accident analysis code MELCOR, version 1.8.2, has been applied for thermal-hydraulic pre-test analysis of the first test of the Phebus FP program (test FPT-0) to study the best test parameters and the applicability of the code. The Phebus FP program is an in-pile test program which has been planned by the French Commissariate a L'Energie Atomique and the Commission of the European Union. The experiments are being conducted by an international collaboration to study the release and transport of fission products (FPs) under conditions assumed to be the most representative of those that would occur in a severe accident. The Phebus FP test apparatus simulates a test bundle of an in-pile section, the circuit including the steam generator U-tubes and the containment. The FPT-0 test was designed to simulate the heat-up and subsequent fuel bundle degradation after a loss of coolant severe accident, using fresh fuel. Two options for fuel degradation models in MELCOR have been applied to fuel degradation behavior. the first model assumes that fuel debris will be formed immediately after the fuel support fails by cladding relocation due to the candling process. The other is the uncollapsed bare fuel pellets option, in which the fuel pellets remain standing in a columnar shape until the fuel reaches its melting point, even if the cladding has been relocated by candling. The thermal-hydraulic behaviors in the circuit and containment of Phebus FP are discussed herein. Flow velocities in the Phebus FP circuit are high in order to produce turbulent flow in a small diameter test pipe. The MELCOR calculation has shown that the length of the hot leg and steam generator are adequate to attain steam temperatures or 700 degrees C and 150 degrees C in the respective outlets. The containment atmosphere temperature and humidity derived by once through integral system calculation show that objective test conditions would be satisfied in the Phebus FP experiment

  7. The Phebus FP thermal-hydraulic analysis with Melcor

    Energy Technology Data Exchange (ETDEWEB)

    Akgane, Kikuo; Kiso, Yoshihiro [Nuclear Power Engineering Corporation, Tokyo (Japan); Fukahori, Takanori [Hitachi Engineering Company, Ltd., Hitachi-shi Ibaraki-ken (Japan); Yoshino, Mamoru [Nuclear Engineering Ltd., Tosabori Nishi-ku (Japan)

    1995-09-01

    The severe accident analysis code MELCOR, version 1.8.2, has been applied for thermal-hydraulic pre-test analysis of the first test of the Phebus FP program (test FPT-0) to study the best test parameters and the applicability of the code. The Phebus FP program is an in-pile test program which has been planned by the French Commissariate a L`Energie Atomique and the Commission of the European Union. The experiments are being conducted by an international collaboration to study the release and transport of fission products (FPs) under conditions assumed to be the most representative of those that would occur in a severe accident. The Phebus FP test apparatus simulates a test bundle of an in-pile section, the circuit including the steam generator U-tubes and the containment. The FPT-0 test was designed to simulate the heat-up and subsequent fuel bundle degradation after a loss of coolant severe accident, using fresh fuel. Two options for fuel degradation models in MELCOR have been applied to fuel degradation behavior. the first model assumes that fuel debris will be formed immediately after the fuel support fails by cladding relocation due to the candling process. The other is the uncollapsed bare fuel pellets option, in which the fuel pellets remain standing in a columnar shape until the fuel reaches its melting point, even if the cladding has been relocated by candling. The thermal-hydraulic behaviors in the circuit and containment of Phebus FP are discussed herein. Flow velocities in the Phebus FP circuit are high in order to produce turbulent flow in a small diameter test pipe. The MELCOR calculation has shown that the length of the hot leg and steam generator are adequate to attain steam temperatures or 700{degrees}C and 150{degrees}C in the respective outlets. The containment atmosphere temperature and humidity derived by once through integral system calculation show that objective test conditions would be satisfied in the Phebus FP experiment.

  8. Transmutation technology development; thermal hydraulic power analysis and structure analysis of the HYPER target beam window

    Energy Technology Data Exchange (ETDEWEB)

    Choi, J. H.; Ju, E. S.; Song, M. K.; Jeon, Y. Z. [Gyeongsang National University, Jinju (Korea)

    2002-03-01

    A thermal hydraulic power analysis, a structure analysis and optimization computation for some design factor for the design of spallation target suitable for HYPER with 1000 MW thermal power in this study was performed. Heat generation formula was used which was evaluated recently based on the LAHET code, mainly to find the maximum beam current under given computation conditions. Thermal hydraulic power of HYPER target system was calculated using FLUENT code, structure conducted by inputting the data into ANSYS. On the temp of beam windows and the pressure distribution calculated using FLUENT. Data transformation program was composed apply the data calculated using FLUENT being commercial CFD code and ANSYS being FEM code for CFX structure analysis. A basic study was conducted on various singular target to obtain fundamental data on the shape for optimum target design. A thermal hydraulic power analysis and structure analysis were conducted on the shapes of parabolic, uniform, scanning beams to choose the optimum shape of beam current analysis was done according to some turbulent model to simulate the real flow. To evaluate the reliability of numerical analysis result, benchmarking of FLUENT code reformed at SNU and Korea Advanced Institute of Science and Technology and it was compared to CFX in the possession of Korea Atomic Energy Research Institute and evaluated. Reliable deviation was observed in the results calculated using FLUENT code, but temperature deviation of about 200 .deg. C was observed in the result from CFX analysis at optimum design condition. Several benchmarking were performed on the basis of numerical analysis concerning conventional HYPER. It was possible to allow a beam arrests of 17.3 mA in the case of the {phi} 350 mm parabolic beam suggested to the optimum in nuclear transmutation when stress equivalent to VON-MISES was calculated to be 140 MPa. 29 refs., 109 figs. (Author)

  9. Design and thermal-hydraulic analysis of PFC baking for SST-1 Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Chaudhuri, Paritosh E-mail: paritosh@ipr.res.in; Reddy, D. Chenna; Khirwadkar, S.; Prakash, N. Ravi; Santra, P.; Saxena, Y.C

    2001-09-01

    The Steady-State Superconducting Tokamak (SST-1) is a medium-size tokamak with super-conducting magnetic field coils. Plasma facing components (PFC) of the SST-1, consisting of divertors, passive stabilisers, baffles, and poloidal limiters, are designed to be compatible for steady-state operation. Except for the poloidal limiters, all other PFC are structurally continuous in the toroidal direction. As SST-1 is designed to run double-null divertor plasmas, these components also have up-down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. The passive stabilisers are located close to the plasma to provide stability against the vertical instability of the elongated plasma. The main consideration in the design of the PFC is the steady-state heat removal of up to 1 MW/m{sup 2}. In addition to removing high heat fluxes, the PFC are also designed to be compatible for baking at 350 deg. C. Different flow parameters and various tube layouts have been examined to select the optimum thermal-hydraulic parameters and tube layout for different PFC of SST-1. Thermal response of the PFC during baking has been performed analytically (using a Fortran code) and two-dimensional finite element analysis using ANSYS. The detailed thermal hydraulics and thermal responses of PFC baking is presented in this paper.

  10. Design and thermal-hydraulic analysis of PFC baking for SST-1 Tokamak

    International Nuclear Information System (INIS)

    Chaudhuri, Paritosh; Reddy, D. Chenna; Khirwadkar, S.; Prakash, N. Ravi; Santra, P.; Saxena, Y.C.

    2001-01-01

    The Steady-State Superconducting Tokamak (SST-1) is a medium-size tokamak with super-conducting magnetic field coils. Plasma facing components (PFC) of the SST-1, consisting of divertors, passive stabilisers, baffles, and poloidal limiters, are designed to be compatible for steady-state operation. Except for the poloidal limiters, all other PFC are structurally continuous in the toroidal direction. As SST-1 is designed to run double-null divertor plasmas, these components also have up-down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. The passive stabilisers are located close to the plasma to provide stability against the vertical instability of the elongated plasma. The main consideration in the design of the PFC is the steady-state heat removal of up to 1 MW/m 2 . In addition to removing high heat fluxes, the PFC are also designed to be compatible for baking at 350 deg. C. Different flow parameters and various tube layouts have been examined to select the optimum thermal-hydraulic parameters and tube layout for different PFC of SST-1. Thermal response of the PFC during baking has been performed analytically (using a Fortran code) and two-dimensional finite element analysis using ANSYS. The detailed thermal hydraulics and thermal responses of PFC baking is presented in this paper

  11. Experimental studies on thermal hydraulic responses for transient operations of the SMART-P

    International Nuclear Information System (INIS)

    Choi, K.Y.; Park, H.S.; Cho, S.; Park, C.K.; Lee, S.J.; Song, C.H.; Chung, M.K.

    2005-01-01

    Full text of publication follows: Thermal hydraulic responses for transient operations of the SMART-P are experimentally investigated by using a integral effect test facility. This test facility (VISTA) has been constructed to simulate the SMART-P, which is a pilot plant of the SMART. The SMART-P is an advanced modular integral type pressurized water reactor (65 MWt) whose major RCS components, such as main coolant pumps, helical-coiled tube bundle steam generators and pressurizers, are contained in a reactor vessel. This integral design approach eliminates the large coolant loop piping, thus eliminates the occurrence of a large break LOCA. Passive Residual Heat Removal System (PRHRS) is installed to prevent overheating and over-pressurization of the primary system during accidental conditions. The PRHRS of the SMART-P removes the core decay heat by natural circulation of the two-phase fluid. The VISTA facility is a full height and 1/96 volume scaled test facility with respect to the SMART-P and will be used to understand the thermal-hydraulic responses following transients and finally to verify the system design of the SMART-P. The experimental data from the VISTA facility will be essential to system designers to resolve open issues relevant to the design of the SMART-P. The full functional control logics are implanted into the VISTA facility to cope with abnormal transients. The core of the facility can be selectively controlled by either a T-control or a T+N control method. The T-control method is a control method to adjust the core power according to the core exit coolant temperature and is designed to be used for high primary coolant flow conditions. On the other hand, the T+N control method is for low primary coolant flow conditions and it uses core exit temperature as well as core power itself as control inputs. The thermal hydraulic responses are carefully investigated according to different core control methods. Several experiments have been performed to

  12. VIPRE-01: A thermal-hydraulic code for reactor cores

    International Nuclear Information System (INIS)

    Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.; Nomura, K.K.

    1989-08-01

    The VIPRE-01 thermal hydraulics code for PWR and BWR analysis has undergone significant modifications and error correction. This manual for the updated code, designated as VIPRE-01 Mod-02, describes improvements that eliminate problems of slow convergence with the drift flux model in transient simulation. To update the VIPRE-01 code and its documentation the drift flux model of two-phase flow was implemented and error corrections developed during VIPRE-01 application were included. The project team modified the existing VIPRE-01 equations into drift flux model equations by developing additional terms. They also developed and implemented corrections for the errors identified during the last four years. They then validated the modified code against standard test data using selected test cases. The project team prepared documentation revisions reflecting code improvements and corrections to replace the corresponding sections in the original VIPRE documents. The revised VIPRE code, designated VIPRE-01 Mod-02, incorporates improvements that eliminate many shortcomings of the previous version. During the validation, the code produced satisfactory output compared with test data. The revised documentation is in the form of binder pages to replace existing pages in three of the original manuals

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

    International Nuclear Information System (INIS)

    Shin, C. W.; No, H. C.

    2015-01-01

    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

  14. Thermal hydraulics model for Sandia's annular core research reactor

    International Nuclear Information System (INIS)

    Rao, Dasari V.; El-Genk, Mohamed S.; Rubio, Reuben A.; Bryson, James W.; Foushee, Fabian C.

    1988-01-01

    A thermal hydraulics model was developed for the Annular Core Research Reactor (ACRR) at Sandia National Laboratories. The coupled mass, momentum and energy equations for the core were solved simultaneously using an explicit forward marching numerical technique. The model predictions of the temperature rise across the central channel of the ACRR core were within ± 10 percent agreement with the in-core temperature measurements. The model was then used to estimate the coolant mass flow rate and the axial distribution of the cladding surface temperature in the central and average channels as functions of the operating power and the water inlet subcooling. Results indicated that subcooled boiling occurs at the cladding surface in the central channels of the ACRR at power levels in excess of 0.5 MW. However, the high heat transfer coefficient due to subcooled boiling causes the cladding temperature along most of the active fuel rod region to be quite uniform and to increase very little with the reactor power. (author)

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

    International Nuclear Information System (INIS)

    Ninokata, Hisashi

    2012-01-01

    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.)

  16. 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)

  17. Parallelization methods study of thermal-hydraulics codes

    International Nuclear Information System (INIS)

    Gaudart, Catherine

    2000-01-01

    The variety of parallelization methods and machines leads to a wide selection for programmers. In this study we suggest, in an industrial context, some solutions from the experience acquired through different parallelization methods. The study is about several scientific codes which simulate a large variety of thermal-hydraulics phenomena. A bibliography on parallelization methods and a first analysis of the codes showed the difficulty of our process on the whole applications to study. Therefore, it would be necessary to identify and extract a representative part of these applications and parallelization methods. The linear solver part of the codes forced itself. On this particular part several parallelization methods had been used. From these developments one could estimate the necessary work for a non initiate programmer to parallelize his application, and the impact of the development constraints. The different methods of parallelization tested are the numerical library PETSc, the parallelizer PAF, the language HPF, the formalism PEI and the communications library MPI and PYM. In order to test several methods on different applications and to follow the constraint of minimization of the modifications in codes, a tool called SPS (Server of Parallel Solvers) had be developed. We propose to describe the different constraints about the optimization of codes in an industrial context, to present the solutions given by the tool SPS, to show the development of the linear solver part with the tested parallelization methods and lastly to compare the results against the imposed criteria. (author) [fr

  18. CFD thermal-hydraulic analysis of a CANDU fuel channel

    International Nuclear Information System (INIS)

    Catana, A.; Prisecaru, I.; Dupleac, D.; Danila, N.

    2009-01-01

    This paper presents the numerical investigation of a CANDU fuel channel using CFD (Computational fluid dynamics) methodology approach. Limited computer power available at Bucharest University POLITEHNICA forced the authors to analyse only segments of fuel channel namely the significant ones: fuel bundle junctions with adjacent segments, fuel bundle spacer planes with adjacent segments, regular segments of fuel bundles. The computer code used is FLUENT. Fuel bundles contained in pressure tubes forms a complex flow domain. The flow is characterized by high turbulence and in some parts of fuel channel also by multi-phase flow. The flow in the fuel channel has been simulated by solving the equations for conservation of mass and momentum. For turbulence modelling the standard k-e model is employed although other turbulence models can be used as well. In this paper we do not consider heat generation and heat transfer capabilities of CFD methods. Since we consider only some relatively short segments of a CANDU fuel channel we can assume, for this starting stage, that heat transfer is not very important for these short segments of fuel channel. The boundary conditions for CFD analysis are provided by system and sub-channel analysis. In this paper the discussion is focused on some flow parameters behaviour at the bundle junction, spacer's plane configuration, etc. In this paper we present results for Standard CANDU 6 Fuel Bundles as a basis for CFD thermal-hydraulic analysis of INR proposed SEU43 and other new nuclear fuels. (authors)

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

    International Nuclear Information System (INIS)

    Olive, J.

    1995-11-01

    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

  20. 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.

  1. 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)

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

    International Nuclear Information System (INIS)

    Maruyama, So; Yamashita, Kiyonobu; Fujimoto, Nozomu; Murata, Isao; Sudo, Yukio; Murakami, Tomoyuki; Fujii, Sadao.

    1993-01-01

    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)

  3. Thermal-hydraulic analysis of research reactor core with different LEU fuel types using RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    El-Sahlamy, Neama M. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt)

    2017-11-15

    In the current work, comparisons between the core performances when using different LEU fuels are done. The fuels tested are UA1{sub X}-A1, U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al fuels with 19.7 % enrichment. Calculations are done using RELAP5 code to evaluate the thermal-hydraulic performance of the IAEA benchmark 10 MW reactor. First, a reassessment of the slow reactivity insertion transient with UA1{sub X}-A1 LEU fuel to compare the results with those reported in the IAEA TECDOC [1]. Then, comparisons between the thermal-hydraulic core performances when using the three LEU fuels are done. The assessment is performed at initial power of 1.0 W. The reactor power is calculated using the RELAP5 point kinetic model. The reactivity feedback, from changes in water density and fuel temperature, is considered for all cases. From the results it is noticed that U{sub 3}Si{sub 2}-Al fuel gives the best fuel performance since it has the minimum value of peak fuel temperature and the minimum peak clad surface temperature, as operating parameters. Also, it gives the maximum value of the Critical Heat Flux Ratio and the lowest tendency to flow instability occurrence.

  4. Coupled neutronics/thermal-hydraulics for analysis of molten salt reactor

    International Nuclear Information System (INIS)

    Guo, Zhangpeng; Zhou, Jianjun; Zhang, Dalin; Chaudri, Khurrum Saleem; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2013-01-01

    Highlights: ► A multiple-channel analysis code (MAC) is developed to be coupled with MCNP. ► 1/8 of core is simulated in MCNP and thermal-hydraulic code. ► The coupling calculation can achieve stable state after a few iterations. ► The coupling calculation results are in reasonable agreement with the analytic solutions of the ORNL. ► Parametric studies of MSR are performed to provide valuable information for future design MSR. -- Abstract: The Generation IV International Forum (GIF) selected molten salt reactor (MSR) among six advanced reactor types. It is characterized by a liquid circulating fuel that also serves as coolant. In this study, a multiple-channel analysis code (MAC) is developed and it is coupled with MCNP4c to analyze the neutronics/thermal-hydraulics behavior of molten salt reactor experiment (MSRE). The MAC calculates thermal-hydraulic parameters, such as temperature distribution, flow distribution and pressure drop. MCNP4c performs the analysis of effective multiplication factor, neutron flux and power distribution. A linkage code is developed to exchange data between MAC and MCNP to implement coupling iteration process until the power convergence is achieved. The coupling calculation can achieve converged solution after a few iterations. The results are in reasonable agreement with the analytic solutions from the ORNL. For further design analysis, parametric studies are performed to provide valuable information for new design of MSR. The effect of inlet temperature, graphite to molten salt volume ratio (G/Ms) from varying channel diameter and different power levels on the effective multiplication factor, neutron flux, graphite lifetime and temperature distribution are discussed in detail

  5. Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

    Energy Technology Data Exchange (ETDEWEB)

    Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro, E-mail: duvan.castellanos@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: pedro.rossi@ufabc.edu.br, E-mail: pedro.carajilescov10@gmail.com [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil). Centro de Engenharias, Modelagem e Ciências Sociais Aplicadas

    2017-07-01

    To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

  6. Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

    International Nuclear Information System (INIS)

    Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro

    2017-01-01

    To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

  7. Coupled neutronics and thermal hydraulics of high density cores for FRM II

    Energy Technology Data Exchange (ETDEWEB)

    Breitkreutz, Harald

    2011-03-04

    According to the 'Verwaltungsvereinbarung zwischen Bund und Land vom 30.5.2003' and its updating on 13.11.2010, the Forschungs-Neutronenquelle Heinz Maier-Leibnitz, Frm II, has to convert its fuel element to an uranium enrichment which is significantly lower than the current 93%, in case this is economically reasonable and doesn't impact the reactor performance immoderate. In the framework of this conversion, new calculations regarding neutronics and thermal hydraulics for the anticipated core configurations have to be made. The computational power available nowadays allows for detailed 3D calculations, on the neutronic as well as on the thermal hydraulic side. In this context, a new program system, 'X{sup 2}', was developed. It couples the Monte Carlo code McnpX, the computational fluid dynamics code Cfx and the burn-up code sequence MonteBurns. The codes were modified and extended to meet the requirements of the coupled calculation concept. To verify the new program system, highly detailed calculations for the current fuel element were made and compared to simulations and measurements that were performed in the past. The results strengthen the works performed so far and show that the original, conservative approach overestimates all critical thermal hydraulic values. Using the CFD software, effects like the impact of the combs that fix the fuel plates and the pressure drop at the edges of the fuel plates were studied in great detail for the first time. Afterwards, a number of possible new fuel elements with lower enrichment, based on disperse and monolithic UMo (uranium with 8 wt.-% Mo) were analysed. A number of straight-forward conversion scenarios was discussed, showing that a further compaction of the fuel element, an extended cycle length or an increased reactor power is needed to compensate the flux loss, which is caused by the lower enrichment. This flux loss is in excess of 7%. The discussed new fuel elements include a 50

  8. Validation studies of thermal-hydraulic code for safety analysis of nuclear power plants

    International Nuclear Information System (INIS)

    Haapalehto, T.

    1995-01-01

    The thesis gives an overview of the validation process for thermal-hydraulic system codes and it presents in more detail the assessment and validation of the French code CATHARE for VVER calculations. Three assessment cases are presented: loop seal clearing, core reflooding and flow in a horizontal steam generator. The experience gained during these assessment and validation calculations has been used to analyze the behavior of the horizontal steam generator and the natural circulation in the geometry of the Loviisa nuclear power plant. Large part of the work has been performed in cooperation with the CATHARE-team in Grenoble, France. (41 refs., 11 figs., 8 tabs.)

  9. Best estimate LB LOCA approach based on advanced thermal-hydraulic codes

    International Nuclear Information System (INIS)

    Sauvage, J.Y.; Gandrille, J.L.; Gaurrand, M.; Rochwerger, D.; Thibaudeau, J.; Viloteau, E.

    2004-01-01

    Improvements achieved in thermal-hydraulics with development of Best Estimate computer codes, have led number of Safety Authorities to preconize realistic analyses instead of conservative calculations. The potentiality of a Best Estimate approach for the analysis of LOCAs urged FRAMATOME to early enter into the development with CEA and EDF of the 2nd generation code CATHARE, then of a LBLOCA BE methodology with BWNT following the Code Scaling Applicability and Uncertainty (CSAU) proceeding. CATHARE and TRAC are the basic tools for LOCA studies which will be performed by FRAMATOME according to either a deterministic better estimate (dbe) methodology or a Statistical Best Estimate (SBE) methodology. (author)

  10. Coupled neutronics and thermal hydraulics of high density cores for FRM II

    International Nuclear Information System (INIS)

    Breitkreutz, Harald

    2011-01-01

    According to the 'Verwaltungsvereinbarung zwischen Bund und Land vom 30.5.2003' and its updating on 13.11.2010, the Forschungs-Neutronenquelle Heinz Maier-Leibnitz, Frm II, has to convert its fuel element to an uranium enrichment which is significantly lower than the current 93%, in case this is economically reasonable and doesn't impact the reactor performance immoderate. In the framework of this conversion, new calculations regarding neutronics and thermal hydraulics for the anticipated core configurations have to be made. The computational power available nowadays allows for detailed 3D calculations, on the neutronic as well as on the thermal hydraulic side. In this context, a new program system, 'X 2 ', was developed. It couples the Monte Carlo code McnpX, the computational fluid dynamics code Cfx and the burn-up code sequence MonteBurns. The codes were modified and extended to meet the requirements of the coupled calculation concept. To verify the new program system, highly detailed calculations for the current fuel element were made and compared to simulations and measurements that were performed in the past. The results strengthen the works performed so far and show that the original, conservative approach overestimates all critical thermal hydraulic values. Using the CFD software, effects like the impact of the combs that fix the fuel plates and the pressure drop at the edges of the fuel plates were studied in great detail for the first time. Afterwards, a number of possible new fuel elements with lower enrichment, based on disperse and monolithic UMo (uranium with 8 wt.-% Mo) were analysed. A number of straight-forward conversion scenarios was discussed, showing that a further compaction of the fuel element, an extended cycle length or an increased reactor power is needed to compensate the flux loss, which is caused by the lower enrichment. This flux loss is in excess of 7%. The discussed new fuel elements include a 50% enriched disperse UMo core with

  11. Proceedings of transient thermal-hydraulics and coupled vessel and piping system responses 1991

    International Nuclear Information System (INIS)

    Wang, G.Y.; Shin, Y.W.; Moody, F.J.

    1991-01-01

    This book reports on transient thermal-hydraulics and coupled vessel and piping system responses. Topics covered include: nuclear power plant containment designs; analysis of control rods; gate closure of hydraulic turbines; and shock wave solutions for steam water mixtures in piping systems

  12. SMITHERS: An object-oriented modular mapping methodology for MCNP-based neutronic–thermal hydraulic multiphysics

    International Nuclear Information System (INIS)

    Richard, Joshua; Galloway, Jack; Fensin, Michael; Trellue, Holly

    2015-01-01

    Highlights: • A modular mapping methodogy for neutronic-thermal hydraulic nuclear reactor multiphysics, SMITHERS, has been developed. • Written in Python, SMITHERS takes a novel object-oriented approach for facilitating data transitions between solvers. This approach enables near-instant compatibility with existing MCNP/MONTEBURNS input decks. • It also allows for coupling with thermal-hydraulic solvers of various levels of fidelity. • Two BWR and PWR test problems are presented for verifying correct functionality of the SMITHERS code routines. - Abstract: A novel object-oriented modular mapping methodology for externally coupled neutronics–thermal hydraulics multiphysics simulations was developed. The Simulator using MCNP with Integrated Thermal-Hydraulics for Exploratory Reactor Studies (SMITHERS) code performs on-the-fly mapping of material-wise power distribution tallies implemented by MCNP-based neutron transport/depletion solvers for use in estimating coolant temperature and density distributions with a separate thermal-hydraulic solver. The key development of SMITHERS is that it reconstructs the hierarchical geometry structure of the material-wise power generation tallies from the depletion solver automatically, with only a modicum of additional information required from the user. Additionally, it performs the basis mapping from the combinatorial geometry of the depletion solver to the required geometry of the thermal-hydraulic solver in a generalizable manner, such that it can transparently accommodate varying levels of thermal-hydraulic solver geometric fidelity, from the nodal geometry of multi-channel analysis solvers to the pin-cell level of discretization for sub-channel analysis solvers. The mapping methodology was specifically developed to be flexible enough such that it could successfully integrate preexisting depletion solver case files with different thermal-hydraulic solvers. This approach allows the user to tailor the selection of a

  13. Overcooling transient selection and thermal hydraulic analyses of the Loviisa PTS assessments

    Energy Technology Data Exchange (ETDEWEB)

    Tuomisto, H [IVO Power Engineering Ltd, Vantaa (Finland)

    1997-09-01

    This paper describes transients selection and thermal hydraulic analyses of various PTS assessment studies performed for the pressure vessels of the Loviisa WWER-reactors. Deterministic analyses have been performed in various stages of the PTS studies and they have always made the formal basis for design and licensing of the reactor pressure vessel. The integrated, probabilistic PTS study was carried out to give an overview of the severity of all different PTS sequences, and give a quantitative estimate of the importance of the PTS issues in relation to the overall safety of the plant. Later, the sequences including external flooding of the pressure vessels were added to the PTS assessment. Thermal recovery annealing of the Loviisa 1 reactor pressure vessel took place during refuelling outage in 1996. (author). 10 refs, 4 figs, 3 tabs.

  14. 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.

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

    International Nuclear Information System (INIS)

    Chung, B. D.; Song, C. H.; Lee, Y. J.; Kwon, T. S.

    2003-04-01

    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. Proceedings of the 10. Meeting on Reactor Physics and Thermal Hydraulics

    International Nuclear Information System (INIS)

    Santos Bastos, W. dos

    1995-01-01

    These proceedings presents all the Meeting papers emphasizing specific aspects on reactor physics method, criticality, fuel management, nuclear data, safety analysis, simulation and shielding, neutronics, thermal hydraulics, reactor operation and computational methods

  17. Computational features of the MELT-III neutronics, thermal-hydraulics computer code system

    International Nuclear Information System (INIS)

    Wilburn, N.P.; Waltar, A.E.

    1976-01-01

    A multichannel, thermal-hydraulics, neutronic accident analysis program for simulating fast reactor behavior from a hypothetical accident inception to the start of core disassembly or to reactor shutdown is described

  18. The Influence of Hydraulic Fracturing on Carbon Storage Performance

    Science.gov (United States)

    Fu, Pengcheng; Settgast, Randolph R.; Hao, Yue; Morris, Joseph P.; Ryerson, Frederick J.

    2017-12-01

    Conventional principles of the design and operation of geologic carbon storage (GCS) require injecting CO2 below the caprock fracturing pressure to ensure the integrity of the storage complex. In nonideal storage reservoirs with relatively low permeability, pressure buildup can lead to hydraulic fracturing of the reservoir and caprock. While the GCS community has generally viewed hydraulic fractures as a key risk to storage integrity, a carefully designed stimulation treatment under appropriate geologic conditions could provide improved injectivity while maintaining overall seal integrity. A vertically contained hydraulic fracture, either in the reservoir rock or extending a limited height into the caprock, provides an effective means to access reservoir volume far from the injection well. Employing a fully coupled numerical model of hydraulic fracturing, solid deformation, and matrix fluid flow, we study the enabling conditions, processes, and mechanisms of hydraulic fracturing during CO2 injection. A hydraulic fracture's pressure-limiting behavior dictates that the near-well fluid pressure is only slightly higher than the fracturing pressure of the rock and is insensitive to injection rate and mechanical properties of the formation. Although a fracture contained solely within the reservoir rock with no caprock penetration, would be an ideal scenario, poroelastic principles dictate that sustaining such a fracture could lead to continuously increasing pressure until the caprock fractures. We also investigate the propagation pattern and injection pressure responses of a hydraulic fracture propagating in a caprock subjected to heterogeneous in situ stress. The results have important implications for the use of hydraulic fracturing as a tool for managing storage performance.

  19. Development of subchannel analysis code MATRA-LMR for KALIMER subassembly thermal-hydraulics

    International Nuclear Information System (INIS)

    Won-Seok Kim; Young-Gyun Kim

    2000-01-01

    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)

  20. ARCADIAR - A New Generation of Coupled Neutronics / Core Thermal- Hydraulics Code System at AREVA NP

    International Nuclear Information System (INIS)

    Curca-Tivig, Florin; Merk, Stephan; Pautz, Andreas; Thareau, Sebastien

    2007-01-01

    Anticipating future needs of our customers and willing to concentrate synergies and competences existing in the company for the benefit of our customers, AREVA NP decided in 2002 to develop the next generation of coupled neutronics/ core thermal-hydraulic (TH) code systems for fuel assembly and core design calculations for both, PWR and BWR applications. The global CONVERGENCE project was born: after a feasibility study of one year (2002) and a conceptual phase of another year (2003), development was started at the beginning of 2004. The present paper introduces the CONVERGENCE project, presents the main feature of the new code system ARCADIA R and concludes on customer benefits. ARCADIA R is designed to meet AREVA NP market and customers' requirements worldwide. Besides state-of-the-art physical modeling, numerical performance and industrial functionality, the ARCADIA R system is featuring state-of-the-art software engineering. The new code system will bring a series of benefits for our customers: e.g. improved accuracy for heterogeneous cores (MOX/ UOX, Gd...), better description of nuclide chains, and access to local neutronics/ thermal-hydraulics and possibly thermal-mechanical information (3D pin by pin full core modeling). ARCADIA is a registered trademark of AREVA NP. (authors)

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

    International Nuclear Information System (INIS)

    Williams, P.T.; Lucas, A.T.

    1995-08-01

    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. Investigation of film boiling thermal hydraulics under FCI conditions. Results of a numerical study

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Dinh, A.T.; Nourgaliev, R.R.; Sehgal, B.R. [Div. of Nuclear Power Safety Royal Inst. of Tech. (RIT), Brinellvaegen 60, 10044 Stockholm (Sweden)

    1998-01-01

    Film boiling on the surface of a high-temperature melt jet or of a melt particle is one of key phenomena governing the physics of fuel-coolant interactions (FCIs) which may occur during the course of a severe accident in a light water reactor (LWR). A number of experimental and analytical studies have been performed, in the past, to address film boiling heat transfer and the accompanying hydrodynamic aspects. Most of the experiments have, however, been performed for temperature and heat flux conditions, which are significantly lower than the prototypic conditions. For ex-vessel FCIs, high liquid subcooling can significantly affect the FCI thermal hydraulics. Presently, there are large uncertainties in predicting natural-convection film boiling of subcooled liquids on high-temperature surfaces. In this paper, research conducted at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS), Stockholm, concerning film-boiling thermal hydraulics under FCI condition is presented. Notably, the focus is placed on the effects of (1) water subcooling, (2) high-temperature steam properties, (3) the radiation heat transfer and (4) mixing zone boiling dynamics, on the vapor film characteristics. Numerical investigations are performed using a novel CFD modeling concept named as the local-homogeneous-slip model (LHSM). Results of the analytical and numerical studies are discussed with respect to boiling dynamics under FCI conditions. (author)

  3. Validation of thermal hydraulic codes for fusion reactors safety

    International Nuclear Information System (INIS)

    Sardain, P.; Gulden, W.; Massaut, V.; Takase, K.; Merill, B.; Caruso, G.

    2006-01-01

    A significant effort has been done worldwide on the validation of thermal hydraulic codes, which can be used for the safety assessment of fusion reactors. This work is an item of an implementing agreement under the umbrella of the International Energy Agency. The European part is supported by EFDA. Several programmes related to transient analysis in water-cooled fusion reactors were run in order to assess the capabilities of the codes to treat the main physical phenomena governing the accidental sequences related to water/steam discharge into the vacuum vessel or the cryostat. The typical phenomena are namely the pressurization of a volume at low initial pressure, the critical flow, the flashing, the relief into an expansion volume, the condensation of vapor in a pressure suppression system, the formation of ice on a cryogenic structure, the heat transfer between walls and fluid in various thermodynamic conditions. · A benchmark exercise has been done involving different types of codes, from homogeneous equilibrium to six equations non-equilibrium models. Several cases were defined, each one focusing on a particular phenomenon. · The ICE (Ingress of Coolant Event) facility has been operated in Japan. It has simulated an in-vessel LOCA and the discharge of steam into a pressure suppression system. · The EVITA (European Vacuum Impingement Test Apparatus) facility has been operated in France. It has simulated ingress of coolant into the cryostat, i.e. into a volume at low initial pressure containing surfaces at cryogenic temperature. This paper gives the main lessons gained from these programs, in particular the possibilities for the improvement of the computer codes, extending their capabilities. For example, the water properties have been extended below the triple point. Ice formation models have been implemented. Work has also been done on condensation models. The remaining needs for R-and-D are also highlighted. (author)

  4. VIPRE-01. a thermal-hydraulic analysis code for reactor cores. Volume 1. Mathematical modeling

    International Nuclear Information System (INIS)

    Stewart, C.W.; Cuta, J.M.; Koontz, A.S.; Kelly, J.M.; Basehore, K.L.; George, T.L.; Rowe, D.S.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 1: Mathematical Modeling) explains the major thermal hydraulic models and supporting correlations in detail

  5. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Komen, E M.J.; Bogaard, J.P.A. van den

    1995-06-01

    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.).

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

    International Nuclear Information System (INIS)

    Meng Lin; Rui Hu; Yun Su; Ronghua Zhang; Yanhua Yang

    2005-01-01

    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

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

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Kukita; Ohnuki, Akira

    1997-01-01

    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

  9. 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).

  10. Resolution of thermal-hydraulic safety and licensing issues for the system 80+trademark design

    International Nuclear Information System (INIS)

    Carpentino, S.E.; Ritterbusch, S.E.; Schneider, R.E.

    1995-01-01

    The System 80+ 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

  11. 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.

  12. A flexible coupling scheme for Monte Carlo and thermal-hydraulics codes

    Energy Technology Data Exchange (ETDEWEB)

    Hoogenboom, J. Eduard, E-mail: J.E.Hoogenboom@tudelft.nl [Delft University of Technology (Netherlands); Ivanov, Aleksandar; Sanchez, Victor, E-mail: Aleksandar.Ivanov@kit.edu, E-mail: Victor.Sanchez@kit.edu [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Eggenstein-Leopoldshafen (Germany); Diop, Cheikh, E-mail: Cheikh.Diop@cea.fr [CEA/DEN/DANS/DM2S/SERMA, Commissariat a l' Energie Atomique, Gif-sur-Yvette (France)

    2011-07-01

    A coupling scheme between a Monte Carlo code and a thermal-hydraulics code is being developed within the European NURISP project for comprehensive and validated reactor analysis. The scheme is flexible as it allows different Monte Carlo codes and different thermal-hydraulics codes to be used. At present the MCNP and TRIPOLI4 Monte Carlo codes can be used and the FLICA4 and SubChanFlow thermal-hydraulics codes. For all these codes only an original executable is necessary. A Python script drives the iterations between Monte Carlo and thermal-hydraulics calculations. It also calls a conversion program to merge a master input file for the Monte Carlo code with the appropriate temperature and coolant density data from the thermal-hydraulics calculation. Likewise it calls another conversion program to merge a master input file for the thermal-hydraulics code with the power distribution data from the Monte Carlo calculation. Special attention is given to the neutron cross section data for the various required temperatures in the Monte Carlo calculation. Results are shown for an infinite lattice of PWR fuel pin cells and a 3 x 3 fuel BWR pin cell cluster. Various possibilities for further improvement and optimization of the coupling system are discussed. (author)

  13. Development of MARS for multi-dimensional and multi-purpose thermal-hydraulic system analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Chung, Bub Dong; Kim, Kyung Doo; Hwang, Moon Kyu; Jeong, Jae Jun; Ha, Kwi Seok; Joo, Han Gyu [Korea Atomic Energy Research Institute, T/H Safety Research Team, Yusung, Daejeon (Korea)

    2000-10-01

    MARS (Multi-dimensional Analysis of Reactor Safety) code is being developed by KAERI for the realistic thermal-hydraulic simulation of light water reactor system transients. MARS 1.4 has been developed as a final version of basic code frame for the multi-dimensional analysis of system thermal-hydraulics. Since MARS 1.3, MARS 1.4 has been improved to have the enhanced code capability and user friendliness through the unification of input/output features, code models and code functions, and through the code modernization. Further improvements of thermal-hydraulic models, numerical method and user friendliness are being carried out for the enhanced code accuracy. As a multi-purpose safety analysis code system, a coupled analysis system, MARS/MASTER/CONTEMPT, has been developed using multiple DLL (Dynamic Link Library) techniques of Windows system. This code system enables the coupled, that is, more realistic analysis of multi-dimensional thermal-hydraulics (MARS 2.0), three-dimensional core kinetics (MASTER) and containment thermal-hydraulics (CONTEMPT). This paper discusses the MARS development program, and the developmental progress of the MARS 1.4 and the MARS/MASTER/CONTEMPT focusing on major features of the codes and their verification. It also discusses thermal hydraulic models and new code features under development. (author)

  14. A flexible coupling scheme for Monte Carlo and thermal-hydraulics codes

    International Nuclear Information System (INIS)

    Hoogenboom, J. Eduard; Ivanov, Aleksandar; Sanchez, Victor; Diop, Cheikh

    2011-01-01

    A coupling scheme between a Monte Carlo code and a thermal-hydraulics code is being developed within the European NURISP project for comprehensive and validated reactor analysis. The scheme is flexible as it allows different Monte Carlo codes and different thermal-hydraulics codes to be used. At present the MCNP and TRIPOLI4 Monte Carlo codes can be used and the FLICA4 and SubChanFlow thermal-hydraulics codes. For all these codes only an original executable is necessary. A Python script drives the iterations between Monte Carlo and thermal-hydraulics calculations. It also calls a conversion program to merge a master input file for the Monte Carlo code with the appropriate temperature and coolant density data from the thermal-hydraulics calculation. Likewise it calls another conversion program to merge a master input file for the thermal-hydraulics code with the power distribution data from the Monte Carlo calculation. Special attention is given to the neutron cross section data for the various required temperatures in the Monte Carlo calculation. Results are shown for an infinite lattice of PWR fuel pin cells and a 3 x 3 fuel BWR pin cell cluster. Various possibilities for further improvement and optimization of the coupling system are discussed. (author)

  15. Development of MARS for multi-dimensional and multi-purpose thermal-hydraulic system analysis

    International Nuclear Information System (INIS)

    Lee, Won Jae; Chung, Bub Dong; Kim, Kyung Doo; Hwang, Moon Kyu; Jeong, Jae Jun; Ha, Kwi Seok; Joo, Han Gyu

    2000-01-01

    MARS (Multi-dimensional Analysis of Reactor Safety) code is being developed by KAERI for the realistic thermal-hydraulic simulation of light water reactor system transients. MARS 1.4 has been developed as a final version of basic code frame for the multi-dimensional analysis of system thermal-hydraulics. Since MARS 1.3, MARS 1.4 has been improved to have the enhanced code capability and user friendliness through the unification of input/output features, code models and code functions, and through the code modernization. Further improvements of thermal-hydraulic models, numerical method and user friendliness are being carried out for the enhanced code accuracy. As a multi-purpose safety analysis code system, a coupled analysis system, MARS/MASTER/CONTEMPT, has been developed using multiple DLL (Dynamic Link Library) techniques of Windows system. This code system enables the coupled, that is, more realistic analysis of multi-dimensional thermal-hydraulics (MARS 2.0), three-dimensional core kinetics (MASTER) and containment thermal-hydraulics (CONTEMPT). This paper discusses the MARS development program, and the developmental progress of the MARS 1.4 and the MARS/MASTER/CONTEMPT focusing on major features of the codes and their verification. It also discusses thermal hydraulic models and new code features under development. (author)

  16. Data report of BWR post-CHF tests. Transient core thermal-hydraulic test program. Contract research

    International Nuclear Information System (INIS)

    Iguchi, Tadashi; Itoh, Hideo; Kiuchi, Toshio; Watanabe, Hironori; Kimura, Mamoru; Anoda, Yoshinari

    2001-03-01

    JAERI has been performing transient core thermal-hydraulic test program. In the program, authors performed BWR/ABWR DBE simulation tests with a test facility, which can simulate BWR/ABWR transients. The test facility has a 4 x 4 bundle core simulator with 15-rod heaters and one non-heated rod. Through the tests, authors quantified the thermal safety margin for core cooling. In order to quantify the thermal safety margin, authors collected experimental data on post-CHF. The data are essential for the evaluation of clad temperature transient when core heat-up occurs during DBEs. In comparison with previous post-CHF tests, present experiments were performed in much wider experimental condition, covering high clad temperature, low to high pressure and low to high mass flux. Further, data at wider elevation (lower to higher elevation of core) were obtained in the present experiments, which make possible to discuss the effect of axial position on thermal-hydraulics, while previous works usually discuss the thermal-hydraulics at the position where the first heat-up occurs. This data report describes test procedure, test condition and major experimental data of post-CHF tests. (author)

  17. Thermal-Hydraulic Sensitivity Study of Intermediate Loop Parameters for Nuclear Hydrogen Production System

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jong Hwa; Lee, Heung Nae; Park, Jea Ho [KONES Corp., Seoul (Korea, Republic of); Lee, Won Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Sang Il; Yoo, Yeon Jae [Hyundai Engineering Co., Seoul (Korea, Republic of)

    2016-10-15

    The heat generated from the VHTR is transferred to the intermediate loop through Intermediate Heat Exchanger (IHX). It is further passed on to the Sulfur-Iodine (SI) hydrogen production system (HPS) through Process Heat Exchanger (PHX). The IL provides the safety distance between the VHTR and HPS. Since the IL performance affects the overall nuclear HPS efficiency, it is required to optimize its design and operation parameters. In this study, the thermal-hydraulic sensitivity of IL parameters with various coolant options has been examined by using MARS-GCR code, which was already applied for the case of steam generator. Sensitivity study of the IL and PHX parameters has been carried out based on their thermal-hydraulic performance. Several parameters for design and operation, such as the pipe diameter, safety distance and surface area, are considered for different coolant options, He, CO{sub 2} and He-CO{sub 2} (2:8). It was found that the circulator work is the major factor affecting on the overall nuclear hydrogen production system efficiency. Circulator work increases with the safety distance, and decreases with the operation pressure and loop pipe diameter. Sensitivity results obtained from this study will contribute to the optimization of the IL design and operation parameters and the optimal coolant selection.

  18. Development of the Real-time Core and Thermal-Hydraulic Models for Kori-1 Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jin Hyuk; Lee, Myeong Soo; Hwang, Do Hyun; Byon, Soo Jin [KEPRI, Daejeon (Korea, Republic of)

    2010-10-15

    The operation of the Kori-Unit 1 (1723.5MWt) is expanded to additional 10 years with upgrades of the Main Control Room (MCR). Therefore, the revision of the procedures, performance tests and works related with the exchange of the Main Control Board (MCB) are currently carried out. And as a part of it, the fullscope simulator for the Kori-1 is being developed for the purpose of the pre-operation and emergence response capability for the operators. The purpose of this paper is to report on the performance of the developed neutronics and thermal-hydraulic (TH) models of Kori Unit 1 simulator. The neutronics model is based on the NESTLE code and TH model based on the RELAP5/MOD3 thermal-hydraulics analysis code which was funded as FY-93 LDRD Project 7201 and is running on the commercial simulator environment tool (the 3KeyMaster{sup TM} of the WSC). As some examples for the verification of the developed neutronics and TH models, some figures are provided. The outputs of the developed neutronics and TH models are in accord with the Nuclear Design Report (NDR) and Final Safety Analysis Report (FSAR) of the reference plant

  19. Steam generator thermal hydraulic design & functional architecture features and related operational and reliability issues requiring consideration

    International Nuclear Information System (INIS)

    Klarner, R.G.

    2012-01-01

    Proper thermal hydraulic design and functional architecture are critical to successful steam generator operation and long term reliability. The evolution of steam generators has been a gradual learning process that has benefited from continuous industry operational experience (OPEX). Inadequate thermal hydraulic design can lead to numerous degradation mechanisms such as excessive deposition, corrosion, flow and level instabilities, fluid-elastic instabilities and tube wear. The functional architecture determines the health of the tube bundle and the other internals during manufacturing, handling and operation. It also determines thermal performance as well as establishing global thermal-hydraulic characteristics such as water level shrink and swell response. This paper discusses the range of operational and reliability issues and relates them to the thermal hydraulic attributes and functional architecture of steam generators (many SG reliability issues are further discussed in other presentations at this conference). In pursuing such issues, the paper focuses on the four major features of the equipment, identifying in each case the goals and requirements such features must meet. Typical approaches and the means by which such requirements are addressed in current equipment are discussed. The four features are: 1. Tubing Material and Tube Bundle Heat Transfer Performance; a. Two materials are in current use – Alloy 690 TT and Alloy 800. Both are good materials with excellent performance records which serve their owners very well (the reliability attributes of Alloy 800 and 690 are discussed in other papers at this conference). Caution is advised in the supply of any material: – material quality is only assured by what is specified to material suppliers in procurement specifications – i.e. - all the knowledge and research in the world assures nothing if its findings are not reflected in procurement requirements. b. Heat transfer performance in addition to being

  20. Validation of the thermal-hydraulic system code ATHLET based on selected pressure drop and void fraction BFBT tests

    Energy Technology Data Exchange (ETDEWEB)

    Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Escalante, Javier Jimenez; Espinoza, Victor Sanchez

    2015-07-15

    Highlights: • Simulation of BFBT-BWR steady-state and transient tests with ATHLET. • Validation of thermal-hydraulic models based on pressure drops and void fraction measurements. • TRACE system code is used for the comparative study. • Predictions result in a good agreement with the experiments. • Discrepancies are smaller or comparable with respect to the measurements uncertainty. - Abstract: Validation and qualification of thermal-hydraulic system codes based on separate effect tests are essential for the reliability of numerical tools when applied to nuclear power plant analyses. To this purpose, the Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is involved in various validation and qualification activities of different CFD, sub-channel and system codes. In this paper, the capabilities of the thermal-hydraulic code ATHLET are assessed based on the experimental results provided within the NUPEC BFBT benchmark related to key Boiling Water Reactors (BWR) phenomena. Void fraction and pressure drops measurements in the BFBT bundle performed under steady-state and transient conditions which are representative for e.g. turbine trip and recirculation pump trip events, are compared with the numerical results of ATHLET. The comparison of code predictions with the BFBT data has shown good agreement given the experimental uncertainty and the results are consistent with the trends obtained with similar thermal-hydraulic codes.

  1. CFD investigation of thermal-hydraulic characteristics in a PBR core using different contact treatments between pebbles

    International Nuclear Information System (INIS)

    Ferng, Y.M.; Lin, K.Y.

    2014-01-01

    Highlights: • It is important to study thermal-hydraulic characteristics in a PBR for a HTGR. • A CFD model is proposed to simulate flow and heat transfer in a segment of pebbles. • Area and point contact treatments for adjacent pebbles are adopted in this study. • Predicted dependences of Nu and friction factor agree with the correlations. - Abstract: A high temperature gas cooled reactor (HTGR) with a pebble bed core (PBR) can be considered as one of the possible energy generation sources in the incoming future due to its inherently safe performance, lower power density, and higher conversion efficiency, etc. It is important to study the thermal-hydraulic characteristics in a PBR for optimum design and safe operation of a HTGR. In this paper, a computational fluid dynamics (CFD) methodology is proposed to investigate the thermal-hydraulic behavior in a segment of pebbles representing the central region of a PBR. Two kinds of contact modeling between adjacent pebbles are adopted, namely area and point contact treatments. The former contact treatment is a geometric approximation modeling. Based on the comparisons of thermal-hydraulic characteristics in the pebbles predicted by both contact treatments, no significant difference is revealed except for the near-wall secondary flow pattern. In addition, compared with the calculated results from the well-known correlations, the present predicted dependence of Nu number and friction factor on the particle Reynolds number shows good agreement qualitatively and quantitatively

  2. Visualization and measurement by image processing of thermal hydraulic phenomena by neutron radiography

    International Nuclear Information System (INIS)

    Takenaka, Nobuyuki

    1996-01-01

    Neutron Radiography was applied to visualization of thermal hydraulic phenomena and measurement was carried out by image processing the visualized images. Since attenuation of thermal neutron rays is high in ordinary liquids like water and organic fluid while it is low in most of metals, liquid flow behaviors can be visualized through a metallic wall by neutron radiography. Measurement of void fraction and flow vector field which is important to study thermal hydraulic phenomena can be carried out by image processing the images obtained by the visualization. Various two-phase and liquid metal flows were visualized by a JRR-3M thermal neutron radiography system in the present study. Multi-dimensional void fraction distributions in two-phase flows and flow vector fields in liquid metals, which are difficult to measure by the other methods, were successfully measured by image processing. It was shown that neutron radiography was efficiently applicable to study thermal hydraulic phenomena. (author)

  3. Full vessel CFD analysis on thermal-hydraulic characteristics of CPR1000 PWR

    International Nuclear Information System (INIS)

    Chao Yanmeng; Yang Lixin; Zhang Mingqian

    2014-01-01

    To obtain flow distributions and thermal-hydraulic properties in a full vessel PWR under limited computation ability and time, a full vessel simulation model of CPR1000 was built based on two simplification methods. One simplified the inner geometry of the control rod guide tubes using equivalent flow area. Another substituted the core by a porous domain to maintain the pressure drop and temperature rise. After the computation, global and localized flow distributions, hydraulic loads of some main assemblies were obtained, as well as other thermal-hydraulic properties. The results indicate the flow distribution in the full vessel is asymmetrical. Therefore it is essential to use the full vessel model to simulate. The calculated thermal-hydraulic characteristics agree well with the operation statistics, providing the reference data for the reactor safety operation. (authors)

  4. High fidelity thermal-hydraulic analysis using CFD and massively parallel computers

    International Nuclear Information System (INIS)

    Weber, D.P.; Wei, T.Y.C.; Brewster, R.A.; Rock, Daniel T.; Rizwan-uddin

    2000-01-01

    Thermal-hydraulic analyses play an important role in design and reload analysis of nuclear power plants. These analyses have historically relied on early generation computational fluid dynamics capabilities, originally developed in the 1960s and 1970s. Over the last twenty years, however, dramatic improvements in both computational fluid dynamics codes in the commercial sector and in computing power have taken place. These developments offer the possibility of performing large scale, high fidelity, core thermal hydraulics analysis. Such analyses will allow a determination of the conservatism employed in traditional design approaches and possibly justify the operation of nuclear power systems at higher powers without compromising safety margins. The objective of this work is to demonstrate such a large scale analysis approach using a state of the art CFD code, STAR-CD, and the computing power of massively parallel computers, provided by IBM. A high fidelity representation of a current generation PWR was analyzed with the STAR-CD CFD code and the results were compared to traditional analyses based on the VIPRE code. Current design methodology typically involves a simplified representation of the assemblies, where a single average pin is used in each assembly to determine the hot assembly from a whole core analysis. After determining this assembly, increased refinement is used in the hot assembly, and possibly some of its neighbors, to refine the analysis for purposes of calculating DNBR. This latter calculation is performed with sub-channel codes such as VIPRE. The modeling simplifications that are used involve the approximate treatment of surrounding assemblies and coarse representation of the hot assembly, where the subchannel is the lowest level of discretization. In the high fidelity analysis performed in this study, both restrictions have been removed. Within the hot assembly, several hundred thousand to several million computational zones have been used, to

  5. Thermal-hydraulic analysis of spent fuel storage systems

    International Nuclear Information System (INIS)

    Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.

    1987-01-01

    This paper describes the COBRA-SFS (Spent Fuel Storage) computer code, which is designed to predict flow and temperature distributions in spent nuclear fuel storage and transportation systems. The decay heat generated by spent fuel in a dry storage cask is removed through a combination of conduction, natural convection, and thermal radiation. One major advantage of COBRA-SFS is that fluid recirculation within the cask is computed directly by solving the mass and momentum conservation equations. In addition, thermal radiation heat transfer is modeled using detailed radiation exchange factors based on quarter-rod segments. The equations governing mass, momentum, and energy conservation for incompressible flows are presented, and the semi-implicit solution method is described. COBRA-SFS predictions are compared to temperature data from a spent fuel storage cask test and the effect of different fill media on the cladding temperature distribution is discussed. The effect of spent fuel consolidation on cask thermal performance is also investigated. 16 refs., 6 figs., 2 tabs

  6. Thermo-hydraulic performance enhancement of solar air heater ...

    African Journals Online (AJOL)

    DR OKE

    Keywords: Solar air heater; Nusselt number; thermal efficiency; multiple arcs with ... loss; and one or two covers of glass or transparent plastic provide resistance to ..... Methods of testing to determine the thermal performance of solar collectors.

  7. Performance and safety of hydraulic turbines

    International Nuclear Information System (INIS)

    Brekke, H

    2010-01-01

    The first part of the paper contains the choice of small turbines for run of the river power plants. Then a discussion is given on the optimization of the performance of different types of large turbines. Finally a discussion on the safety and necessary maintenance of turbines is given with special attention to bolt connections.

  8. Thermal-hydraulic analysis of the Three Mile Island Unit 2 reactor accident with THALES code

    International Nuclear Information System (INIS)

    Hashimoto, Kazuichiro; Soda, Kunihisa

    1991-10-01

    The OECD Nuclear Energy Agency (NEA) has established a Task Group in the Committee on the Safety of Nuclear Installations (CSNI) to perform an analysis of Three Mile Island Unit 2 (TMI-2) accident as a standard problem to benchmark severe accident computer codes and to assess the capability of the codes. The TMI-2 Analysis Exercise was performed at the Japan Atomic Energy Research Institute (JAERI) using the THALES (Thermal-Hydraulic Analysis of Loss-of-Coolant, Emergency Core Cooling and Severe Core Damage) - PM1/TMI code. The purpose of the analysis is to verify the capability of THALES-PM1/TMI code to describe accident progression in the actual plant. The present paper describes the final result of the TMI-2 Analysis Exercise performed at JAERI. (author)

  9. Development of the NSSS thermal-hydraulic program for YGN unit 1 simulator

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Doo; Jeong, Jae Jun; Lee, Won Jae; Chung, Bub Dong; Ha, Kwi Seok; Kang, Kyung Ho

    2000-09-01

    The NSSS thermal-hydraulic programs installed in the domestic full-scope power plant simulators were provided in early 1980s by foreign vendors. Because of limited computational capability at that time, they usually adopt very simplified physical models for a real-time simulation of NSSS thermal-hydraulic phenomena, which entails inaccurate results and the possibility of so-called 'negative training', especially for complicated two-phase flows in the reactor coolant system. To resolve the problem, we developed a realistic NSSS T/H program (named 'ARTS' code) for use in YongGwang Nuclear Unit 1 full-scope simulator. The best-estimate code RETRAN03, developed by EPRI and approved by USNRC, was selected as a reference code of ARTS. For the development of ARTS, the followings have been performed: -Improvement of the robustness of RETRAN - Improvement of the real-time simulation capability of RETRAN - Optimum input data generation for the NSSS simulation - New model development that cannot be efficiently modeled by RETRAN - Assessment of the ARTS code. The systematic assessment of ARTS has been conducted in both personal computers (Windows 98, Visual fortran) and the simulator development environment (Windows NT, GSE simulator development tool). The results were resonable in terms of accuracy, real-time simulation and robustness.

  10. Parametric analyses of DEMO Divertor using two dimensional transient thermal hydraulic modelling

    Science.gov (United States)

    Domalapally, Phani; Di Caro, Marco

    2018-05-01

    Among the options considered for cooling of the Plasma facing components of the DEMO reactor, water cooling is a conservative option because of its high heat removal capability. In this work a two-dimensional transient thermal hydraulic code is developed to support the design of the divertor for the projected DEMO reactor with water as a coolant. The mathematical model accounts for transient 2D heat conduction in the divertor section. Temperature-dependent properties are used for more accurate analysis. Correlations for single phase flow forced convection, partially developed subcooled nucleate boiling, fully developed subcooled nucleate boiling and film boiling are used to calculate the heat transfer coefficients on the channel side considering the swirl flow, wherein different correlations found in the literature are compared against each other. Correlation for the Critical Heat Flux is used to estimate its limit for a given flow conditions. This paper then investigates the results of the parametric analysis performed, whereby flow velocity, diameter of the coolant channel, thickness of the coolant pipe, thickness of the armor material, inlet temperature and operating pressure affect the behavior of the divertor under steady or transient heat fluxes. This code will help in understanding the basic parameterś effect on the behavior of the divertor, to achieve a better design from a thermal hydraulic point of view.

  11. Sensitivity analysis of thermal hydraulic response in containment at core meltdown accident

    International Nuclear Information System (INIS)

    Kobayashi, Kensuke; Ishigami, Tsutomu; Horii, Hideo; Chiba, Takemi.

    1985-01-01

    A sensitivity analysis of thermal hydraulic response in a containment during a 'station blackout' (the loss of all AC power) accident at Browns Ferry unit one plant was performed with the computer code MARCH 1.0. In the analysis, the plant station batteries were assumed to be available for 4h after the initiation of the accident. The thermal hydraulic response in the containment was calculated by varying several input data for MARCH 1.0 independently and the deviation among calculated results were investigated. The sensitivity analysis showed that (a) the containment would fail due to the overtemperature without any operator actions for plant recovery, which would be strongly dependent on the model of the debris-concrete interaction and the input parameters for specifying the containment failure modes in MARCH 1.0, (b) a core melting temperature and an amount of water left in a primary system at the end of the meltdown were identified as important parameters which influenced the time of the containment failure, and (c) experimental works regarding the parameters mentioned above could be recommended. (author)

  12. Analysis of Thermal-Hydraulic Behavior of CMT in the SMART-ITL Facility

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Byong Guk; Bae, Hwang; Ryu, Sung-Uk; Ryu, Hyobong; Byun, Sun-Joon; Yi, Sung-Jae; Park, Hyun-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    SMART, an integral small modular reactor, received a standard design approval in 2012 and now extends its safety features through replacing active safety injection pumps by passive safety injection systems: core makeup tanks (CMT) and safety injection tanks (SIT). SMART-ITL has been built in a full height scale and 1/49 area and power scale. One train of CMT and SIT has been installed and their thermal-hydraulic behaviors have been identified through a series of tests. In this paper, initial condensation characteristics as well as force balance around the CMT will be discussed for a representative test. PSIS are added into SMART for better treatment of accidents with prolonged station blackout. In the SMART-ITL, the CMT and SIT are installed to evaluate their performance and a series of tests have been conducted. In this paper, the thermal-hydraulic behavior of CMT is addressed based on the experimental data, especially focusing on the issues of fierce condensation after opening of the isolation valve and driving force balance around the CMT.

  13. Regulatory support activities of JNES by thermal-hydraulic and safety analyses

    International Nuclear Information System (INIS)

    Kasahara, Fumio

    2008-01-01

    Current status and some related topics on regulatory support activities of Japan Nuclear Energy Safety Organization (JNES) by thermal-hydraulic and safety analyses are reported. The safety of nuclear facilities is secured primarily by plant owners and operators. However, the regulatory body NISA (Nuclear and Industrial Safety Agency) has conducted a strict regulation to confirm the adequacy of the site condition as well as the basic and detailed design. The JNES has been conducting independent analyses from applicants (audit analyses, etc.) by direction of NISA and supporting its review. In addition to the audit analysis, thermal-hydraulic and safety analyses are used in such areas as analytical evaluation for investigation of causes of accidents and troubles, level 2 PSA for risk informed regulation, etc. Recent activities of audit analyses are for the application of Tsuruga 3 and 4 (APWR), the spent fuel storage facility for the establishment, and the LMFBR Monju for the core change. For the trouble event evaluation, the criticality accident analysis of Sika1 was carried out and the evaluation of effectiveness of accident management (AM) measure for Tomari 3 (PWR) and Monju was performed. The analytical codes for these evaluations are continuously revised by reflecting the state-of-art technical information and validated using the information provided by the data from JAEA, OECD project, etc. (author)

  14. 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)

  15. The Thermal Hydraulics of Tube Support Fouling in Nuclear Steam Generators

    International Nuclear Information System (INIS)

    Rummens, Helena E.C.; Rogers, J.T.; Turner, C.W.

    2004-01-01

    It is hypothesized that the thermal-hydraulic environment plays a role in the fouling of tube supports in nuclear steam generators. Experiments were performed to simulate the thermal-hydraulic environment near various designs of supports. Pressure loss, local velocity, turbulence intensity, and local void fraction were measured to characterize the effect of the support. Fouling mechanisms specific to supports were inferred from these experimental data and from actual steam generator inspection results. An analytical model was developed to predict the rate of particulate deposition on the supports, to better understand the complex processes involved.This paper presents the following set of tools for assessing the fouling propensity of a given support design: (1) proposed fouling mechanisms, (2) criteria for support fouling propensity, (3) correlation of fouling with parameters such as mass flux and quality, (4) descriptions of experimental tools such as flow visualization and measurement of pressure-loss profiles, and (5) analytical tools.An important conclusion from this and our previous work is that the fouling propensity is greater with broached support plates, both trefoil and quatrefoil, than with lattice bar supports and formed bar supports, in which significant cross flows occur

  16. Development of the NSSS thermal-hydraulic program for YGN unit 1 simulator

    International Nuclear Information System (INIS)

    Kim, Kyung Doo; Jeong, Jae Jun; Lee, Won Jae; Chung, Bub Dong; Ha, Kwi Seok; Kang, Kyung Ho

    2000-09-01

    The NSSS thermal-hydraulic programs installed in the domestic full-scope power plant simulators were provided in early 1980s by foreign vendors. Because of limited computational capability at that time, they usually adopt very simplified physical models for a real-time simulation of NSSS thermal-hydraulic phenomena, which entails inaccurate results and the possibility of so-called 'negative training', especially for complicated two-phase flows in the reactor coolant system. To resolve the problem, we developed a realistic NSSS T/H program (named 'ARTS' code) for use in YongGwang Nuclear Unit 1 full-scope simulator. The best-estimate code RETRAN03, developed by EPRI and approved by USNRC, was selected as a reference code of ARTS. For the development of ARTS, the followings have been performed: -Improvement of the robustness of RETRAN - Improvement of the real-time simulation capability of RETRAN - Optimum input data generation for the NSSS simulation - New model development that cannot be efficiently modeled by RETRAN - Assessment of the ARTS code. The systematic assessment of ARTS has been conducted in both personal computers (Windows 98, Visual fortran) and the simulator development environment (Windows NT, GSE simulator development tool). The results were resonable in terms of accuracy, real-time simulation and robustness

  17. 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)

    1998-12-31

    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)

  18. Coupled neutronic/thermal-hydraulic analysis of the HPLWR three pass core

    International Nuclear Information System (INIS)

    Monti, Lanfranco; Starflinger, Joerg; Schulenberg, Thomas

    2008-01-01

    The High Performance Light Water Reactor is an innovative Gen-IV reactor cooled and moderated with water at supercritical pressure. The three pass core concept has been proposed to reduce peaking factors, i.e. hot-channel effects, and it further increases the core heterogeneity, which is mainly due to pronounced water density reduction. For this kind of nuclear reactor, the significant feedbacks - which exist between the properties of the components and the power generation rate - can not be neglected and require a coupled Neutronic/Thermal-Hydraulic analysis even for steady state conditions. The main goal of this paper is to present the developed tool for coupled analyses of the HPLWR. Two state-of-the-art codes have been chosen for Thermal-Hydraulic and Neutronic core analyses, namely TRACE and ERANOS, and they have been coupled with in an iterative procedure in which they are run in series until a steady state condition has been reached. In the simplifying assumptions of uniform enrichment distribution, zero burn-up and ignoring the effect of the control rods, the obtained steady state condition will be discussed and a core power map, flow rate redistribution as well as water and fuel temperature variations will be presented. (author)

  19. Thermal-hydraulic modeling of flow inversion in a research reactor

    International Nuclear Information System (INIS)

    Kazeminejad, H.

    2008-01-01

    The course of loss of flow accident and flow inversion in a pool type research reactor, with scram enabled under natural circulation condition is numerically investigated. The analyses were performed by a lumped parameters approach for the coupled kinetic-thermal-hydraulics, with continuous feedback due to coolant and fuel temperature effects. A modified Runge-Kutta method was adopted for a better solution to the set of stiff differential equations. Transient thermal-hydraulics during the process of flow inversion and establishment of natural circulation were considered for a 10-MW IAEA research reactor. Some important parameters such as the peak temperatures for the hot channel were obtained for both high-enriched and low enriched fuel. The model prediction is also verified through comparison with other computer code results reported in the literature for detailed simulations of loss of flow accidents (LOFA) and the agreement between the results for the peak clad temperatures and key parameters has been satisfactory. It was found that the flow inversion and subsequent establishment of natural circulation keep the peak cladding surface temperature below the saturation temperature to avoid the escalation of clad temperature to the level of onset of nucleate boiling and sub-cooled void formation to ensure the safe operation of the reactor

  20. Analysis of Thermal-Hydraulic Behavior of CMT in the SMART-ITL Facility

    International Nuclear Information System (INIS)

    Jeon, Byong Guk; Bae, Hwang; Ryu, Sung-Uk; Ryu, Hyobong; Byun, Sun-Joon; Yi, Sung-Jae; Park, Hyun-Sik

    2015-01-01

    SMART, an integral small modular reactor, received a standard design approval in 2012 and now extends its safety features through replacing active safety injection pumps by passive safety injection systems: core makeup tanks (CMT) and safety injection tanks (SIT). SMART-ITL has been built in a full height scale and 1/49 area and power scale. One train of CMT and SIT has been installed and their thermal-hydraulic behaviors have been identified through a series of tests. In this paper, initial condensation characteristics as well as force balance around the CMT will be discussed for a representative test. PSIS are added into SMART for better treatment of accidents with prolonged station blackout. In the SMART-ITL, the CMT and SIT are installed to evaluate their performance and a series of tests have been conducted. In this paper, the thermal-hydraulic behavior of CMT is addressed based on the experimental data, especially focusing on the issues of fierce condensation after opening of the isolation valve and driving force balance around the CMT

  1. VUJE's experience in the field of thermal-hydraulic behaviour of WWER

    International Nuclear Information System (INIS)

    Klepach, J.

    1995-01-01

    The thermal-hydraulic behavior (THB) of NPP coolant system and its consequences to nuclear safety of WWER reactors in previous Czechoslovakia has been studied in the VUJE (Nuclear Power Plants Research Institute, Trnava, SK). The institute takes part in the development and verification of its own (SLAP, LENKA, PUMKO, SICHTA, TRACO etc.) and international (DYNAMIKA5) codes for thermal-hydraulic analysis. The verification efforts are concentrated on the WWER specific features such as horizontal steam generators, control and safety system functioning, etc. The whole range of NPP accident analyses is covered by the VUJe staff. The author outlined briefly the WWER specific features as design and implemented improvements in Bohunice V-1 and Mochovce V-1 (WWER 230 model). The pros and cons of the WWER design compared against western type PWR are described. It is believed that although the WWERs are designed under the rules and standards of 1960s, their safety and operational performance can be improved to acceptable level by thorough analysis and appropriate measures. 5 figs

  2. Program ELM: A tool for rapid thermal-hydraulic analysis of solid-core nuclear rocket fuel elements

    International Nuclear Information System (INIS)

    Walton, J.T.

    1992-11-01

    This report reviews the state of the art of thermal-hydraulic analysis codes and presents a new code, Program ELM, for analysis of fuel elements. ELM is a concise computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in a nuclear thermal rocket reactor with axial coolant passages. The program was developed as a tool to swiftly evaluate various heat transfer coefficient and friction factor correlations generated for turbulent pipe flow with heat addition which have been used in previous programs. Thus, a consistent comparison of these correlations was performed, as well as a comparison with data from the NRX reactor experiments from the Nuclear Engine for Rocket Vehicle Applications (NERVA) project. This report describes the ELM Program algorithm, input/output, and validation efforts and provides a listing of the code

  3. FY 1995 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.; McDuffee, J.L.; McFee, M.T.; Ruggles, A.E.; Wendel, M.W.; Yoder, G.L.

    1997-07-01

    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). 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.

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

    International Nuclear Information System (INIS)

    Siman-Tov, M.; Felde, D.K.; Farquharson, G.; McDuffee, J.L.; McFee, M.T.; Ruggles, A.E.; Wendel, M.W.; Yoder, G.L.

    1997-07-01

    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). 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

  5. 2-D CFD time-dependent thermal-hydraulic simulations of CANDU-6 moderator flows

    Energy Technology Data Exchange (ETDEWEB)

    Mehdi Zadeh, Foad [Department of Engineering Physics/Polytechnique Montréal, Montréal, QC (Canada); Étienne, Stéphane [Department of Mechanical Engineering/Polytechnique Montréal, Montréal, QC (Canada); Teyssedou, Alberto, E-mail: alberto.teyssedou@polymtl.ca [Department of Engineering Physics/Polytechnique Montréal, Montréal, QC (Canada)

    2016-12-01

    Highlights: • 2-D time-dependent CFD simulations of CANDU-6 moderator flows are presented. • A thermal-hydraulic code using thermal physical fluid properties is used. • The numerical approach and convergence is validated against available data. • Flow configurations are correlated using Richardson’s number. • Frequency components indicate moderator flow oscillations vs. Richardson numbers. - Abstract: The distribution of the fluid temperature and mass density of the moderator flow in CANDU-6 nuclear power reactors may affect the reactivity coefficient. For this reason, any possible moderator flow configuration and consequently the corresponding temperature distributions must be studied. In particular, the variations of the reactivity may result in major safety issues. For instance, excessive temperature excursions in the vicinity of the calandria tubes nearby local flow stagnation zones, may bring about partial boiling. Moreover, steady-state simulations have shown that for operating condition, intense buoyancy forces may be dominant, which can trigger a thermal stratification. Therefore, the numerical study of the time-dependent flow transition to such a condition, is of fundamental safety concern. Within this framework, this paper presents detailed time-dependent numerical simulations of CANDU-6 moderator flow for a wide range of flow conditions. To get a better insight of the thermal-hydraulic phenomena, the simulations were performed by covering long physical-time periods using an open-source code (Code-Saturne V3) developed by Électricité de France. The results show not only a region where the flow is characterized by coherent structures of flow fluctuations but also the existence of two limit cases where fluid oscillations disappear almost completely.

  6. Development of nuclear transmutation technology - A study on the thermal-hydraulic characteristics of Pb-Bi coolant material

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hyun; You, Young Woo; Cho, Jae Seon; Kim, Ju Youl; Yang, Hui Chang; Huh, Byung Gil [Seoul National University, Seoul (Korea)

    2000-03-01

    The objective of this study is to provide the direction of HYPER design in terms of thermal hydraulics especially through the analysis of thermal hydraulic characteristics of lead-bismuth material as a HYPER coolant and of proton accelerator target system. In this study, in order to evaluate the thermal-hydraulic characteristics of HYPER system, the FLUENT calculation is performed with liquid metal lead-bismuth(43%) and the turbulent Prandtl number model is developed. Also, the heat transfer analyses including temperature rising are performed for accelerator beam window, solid tungsten target and liquid target which is composed of liquid lead and lead-bismuth, respectively and the thermal stress analyses are performed for accelerator beam window. Through this study, the BASECASE whose parameter is HYPER system design specification is calculated by FLUENT. It is shown that the coolant velocity must exceeds 1.6 m/s for supporting the core coolant temperature in operating temperature range. The suggested turbulent Prandtl number model is applicable to liquid metal. And in order to maintain the integrity of proton beam target system, it is necessary to investigate the target structure associated with smoothing the flow path and beam window cooling. 43 refs., 67 figs., 27 tabs. (Author)

  7. Quality assurance of PTS thermal hydraulic calculations at BNL

    International Nuclear Information System (INIS)

    Rohatgi, U.S.; Pu, J.; Jo, J.; Saha, P.

    1983-01-01

    Rapid cooling of the reactor pressure vessel at high pressure has a potential of challenging the vessel integrity. This phenomenon is called overcooling or Pressurized Thermal Shock (PTS). The Nuclear Regulatory Commission (NRC) has selected three plants representing three types of PWRs in use for detailed PTS study. Oconee-1 (B and W), Calvert Cliffs (C.E.), and H.B. Robinson (Westinghouse). The Brookhaven National Laboratory (BNL) has been requested by NRC to review and compare the input decks developed at LANL and INEL, and to compare and explain the differences between the common calculations performed at these two laboratories. However, for the transients that will be computed by only one laboratory, a consistency check will be performed. So far only Oconee-1 calculations have been reviewed at BNL, and the results are presented here

  8. FFTF thermal-hydraulic testing results affecting piping and vessel component design in LMFBR's

    International Nuclear Information System (INIS)

    Stover, R.L.; Beaver, T.R.; Chang, S.C.

    1983-01-01

    The Fast Flux Test Facility completed four years of pre-operational testing in April 1982. This paper describes thermal-hydraulic testing results from this period which impact piping and vessel component design in LMFBRs. Data discussed are piping flow oscillations, piping thermal stratification and vessel upper plenum stratification. Results from testing verified that plant design limits were met

  9. COBRA-3M: a digital computer code for analyzing thermal-hydraulic behavior in pin bundles

    International Nuclear Information System (INIS)

    Marr, W.W.

    1975-03-01

    The COBRA-3M computer program is a modification of the thermal-hydraulic subchannel-analysis program COBRA-III. It includes detailed thermal models of fuel pin and duct wall. It is especially suitable for analyzing small pin bundles used in in-reactor or out-of-reactor experiments. (U.S.)

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

    International Nuclear Information System (INIS)

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

    2007-07-01

    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

  11. Parametric study on thermal-hydraulic characteristics of high conversion light water reactor

    International Nuclear Information System (INIS)

    Mori, Takamasa; Nakagawa, Masayuki; Fujii, Sadao.

    1988-11-01

    To assess the feasibility of high conversion light water reactors (HCLWRs) from the thermal-hydraulic viewpoint, parametric study on thermal-hydraulic characteristics of HCLWR has been carried out by using a unit cell model. It is assumed that a HCLWR core is contained in a current 1000 MWe PWR plant. At the present study, reactor core parameters such as fuel pin diameter, pitch, core height and linear heat rate are widely and parametrically changed to survey the relation between these parameters and the basic thermal-hydraulic characteristics, i.e. maximum fuel temperature, minimum DNBR, reduction of reactor thermal output and so on. The validity of the unit cell model used has been ensured by comparison with the result of a subchannel analysis carried out for a whole core. (author)

  12. Feasibility study for objective oriented design of system thermal hydraulic analysis program

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Jeong, Jae Jun; Hwang, Moon Kyu

    2008-01-01

    The system safety analysis code, such as RELAP5, TRAC, CATHARE etc. have been developed based on Fortran language during the past few decades. Refactoring of conventional codes has been also performed to improve code readability and maintenance. However the programming paradigm in software technology has been changed to use objects oriented programming (OOP), which is based on several techniques, including encapsulation, modularity, polymorphism, and inheritance. In this work, objective oriented program for system safety analysis code has been tried utilizing modernized C language. The analysis, design, implementation and verification steps for OOP system code development are described with some implementation examples. The system code SYSTF based on three-fluid thermal hydraulic solver has been developed by OOP design. The verifications of feasibility are performed with simple fundamental problems and plant models. (author)

  13. 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)

    2000-03-15

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model if existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA analysis. There are three main area of model development, i.e. moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version.

  14. Development of a 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)

    2000-03-01

    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 a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA Analysis. There are three main area of model development, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version. 15 refs., 37 figs., 8 tabs. (Author)

  15. Experiments and analyses in support of the US ALMR thermal-hydraulic design

    International Nuclear Information System (INIS)

    Hunsbedt, A.

    1993-01-01

    The U.S. Advanced Liquid Metal Reactor (ALMR) which is based on the modular PRISM concept utilizes passive safety characteristics to simplify the reactor design and enhance its safety performance. The relatively small size of each reactor facilitates the use of strong negative feedback with rising temperature for inherent reactivity control and direct, natural air cooling for decay heat removal. The tall, slender reactor geometry of the ALMR enhances uniformity and stability of internal flow distribution during steady state operation and natural circulation flow during transient conditions. The flow uniformity and low operating pressure and temperature of the reactor contributes to high structural margins. A number of experiments and associated analyses have been performed to evaluate natural convection and thermal-hydraulic phenomena experienced under decay heat removal conditions. This paper summarizes these various efforts as described separately below and presents the main results. (author)

  16. Testing of Local Velocity Transducer Used at Sodium Thermal Hydraulic Test Facilities

    International Nuclear Information System (INIS)

    Kim, Tae Joon; Eoh, Jae Hyuk; Hwang, In Koo; Jeong, Ji Young; Kim, Jong Man; Lee, Yong Bum; Kim, Yeong Il

    2012-01-01

    KAERI (Korea Atomic Energy Research Institute) will perform a test for a thermal hydraulic simulation with STELLA-1 for a Component Performance Test Sodium Loop in the year 2012, and subsequently it will construct for STELLA-2 for a Sodium Thermalhydraulic Experimental Facility in the year 2016. The STELLA-2 consists of a scaled reactor vessel with a core of electric heaters, four IHXs, two PHTS pumps, two DHXs, and two AHXs. In STELLA-2, several kinds of flow measurements exists. In this paper, the local velocity transducer as a prototype tested in IPPE (in Russia), was manufactured as a prototype by a shop in KAERI. This local velocity transducer will be used to measure the flow rate in a pool

  17. Challenges in thermal and hydraulic analysis of ADS target systems

    International Nuclear Information System (INIS)

    Groetzbach, G.; Batta, A.; Lefhalm, C.-H.; Otic, I.

    2004-01-01

    The liquid metal cooled spallation targets of Accelerator Driven nuclear reactor Systems obey high thermal loads; in addition some flow and cooling conditions are of a prototypical character; in contrast the operating conditions for the engaged materials are narrow; thus, the target development requires a very careful analysis by experimental and numerical means. Especially the cooling of the steel window, which is heated by the proton beam, needs special care. Some of the main goals of the experimental and numerical analyses of the thermal dynamics of those systems are discusses. The prediction of locally detached flows and of flows with larger recirculation areas suffers from insufficient turbulence modeling; this has to be compensated by using prototypical model experiments, e.g. with water, to select the adequate models and numerical schemes. The well known problems with the Reynolds analogy in predicting the heat transfer in liquid metals requires always prototypic liquid metal experiments to select and adapt the turbulent heat flux models. The uncertainties in liquid metal experiments cannot be neglected; so it is necessary to perform CFD calculations and experiments always hand in hand and to develop improve turbulent heat flux models. One contribution to an improved 3 or 4-equation model is deduced from recent Direct Numerical Simulation (DNS) data. (author)

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

    International Nuclear Information System (INIS)

    Roelofs, F.; Batta, A.; Bandini, G.; Van Tichelen, K.; Gerschenfeld, A.; Cheng, X.

    2014-01-01

    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)

  19. Thermal-hydraulic process for cooling, heating and power production with low-grade heat sources in residential sector

    International Nuclear Information System (INIS)

    Borgogno, R.; Mauran, S.; Stitou, D.; Marck, G.

    2017-01-01

    Highlights: • Assessment of solar thermal-hydraulic process for tri-generation application. • Choice of the most suitable working fluid pair (R1234yf/R1233zd). • Evaluation of the global annual performance in Mediterranean climate. • Global annual COP and heat amplification achieving 0.24 and 1.2 respectively. • Global annual performance achieving an electric efficiency of 3.7%. - Abstract: A new process based on thermal-hydraulic conversion actuated by low-grade thermal energy is investigated. Input thermal energy can be provided by the means of solar collectors, as well as other low temperature energy sources. In the following article, “thermo-hydraulic” term refers to a process involving an incompressible fluid used as an intermediate medium to transfer work hydraulically between different thermal operated components or sub-systems. The system aims at providing trigeneration energy features for the residential sector, that is providing heating, cooling and electrical power for meeting the energy needs of domestic houses. This innovative system is made of two dithermal processes (working at two different levels of temperatures) and featuring two different working fluids. The first process is able to directly supply either electrical energy generated by an hydraulic turbine or drives the second process thanks to the incompressible fluid, which is similar to a heat pump effect for heating or cooling purposes. The innovative aspect of this process relies on the use of an hydraulic transfer fluid to transfer the work between each sub-system and therefore simplifying the conversion chain. A model, assuming steady-state operation, is developed to assess the energy performances of different variants of this thermo-hydraulic process with various heat source temperatures (80–110 °C) or heat sinks (0–30 °C), as well as various pairs of working fluids. For instance, in the frame of a single-family home, located in the Mediterranean region, the working

  20. Thermal modeling of a hydraulic hybrid vehicle transmission based on thermodynamic analysis

    International Nuclear Information System (INIS)

    Kwon, Hyukjoon; Sprengel, Michael; Ivantysynova, Monika

    2016-01-01

    Hybrid vehicles have become a popular alternative to conventional powertrain architectures by offering improved fuel efficiency along with a range of environmental benefits. Hydraulic Hybrid Vehicles (HHV) offer one approach to hybridization with many benefits over competing technologies. Among these benefits are lower component costs, more environmentally friendly construction materials, and the ability to recover a greater quantity of energy during regenerative braking which make HHVs partially well suited to urban environments. In order to further the knowledge base regarding HHVs, this paper explores the thermodynamic characteristics of such a system. A system model is detailed for both the hydraulic and thermal components of a closed circuit hydraulic hybrid transmission following the FTP-72 driving cycle. Among the new techniques proposed in this paper is a novel method for capturing rapid thermal transients. This paper concludes by comparing the results of this model with experimental data gathered on a Hardware-in-the-Loop (HIL) transmission dynamometer possessing the same architecture, components, and driving cycle used within the simulation model. This approach can be used for several applications such as thermal stability analysis of HHVs, optimal thermal management, and analysis of the system's thermodynamic efficiency. - Highlights: • Thermal modeling for HHVs is introduced. • A model for the hydraulic and thermal system is developed for HHVs. • A novel method for capturing rapid thermal transients is proposed. • The thermodynamic system diagram of a series HHV is predicted.

  1. The status of thermal-hydraulic studies on the decay heat removal by natural convection using RAMONA and NEPTUN models

    International Nuclear Information System (INIS)

    Hoffmann, H.; Hain, K.; Marten, K.; Rust, K.; Weinberg, D.; Ohira, H.

    2004-01-01

    Thermal-hydraulic experiments were performed with water in order to simulate the decay heat removal by natural convection in a pool-type sodium-cooled reactor. Two test rigs of different scales were used, namely RAMONA (1:20) and NEPTUN (1:5). RAMONA served to study the transition from nominal operation by forced convection to decay heat removal operation by natural convection. Steady-state similarity tests were carried out in both facilities. The investigations cover nominal and non-nominal operation conditions. These data provide a broad basis for the verification of computer programs. Numerical analyses performed with the three-dimensional FLUTAN code indicated that the thermal-hydraulic processes can be quantitatively simulated even for the very complex geometry of the NEPTUN test rig. (author)

  2. Thermal-hydraulics Analysis of a Radioisotope-powered Mars Hopper Propulsion System

    International Nuclear Information System (INIS)

    O'Brien, Robert C.; Klein, Andrew C.; Taitano, William T.; Gibson, Justice; Myers, Brian; Howe, Steven D.

    2011-01-01

    Thermal-hydraulics analyses results produced using a combined suite of computational design and analysis codes are presented for the preliminary design of a concept Radioisotope Thermal Rocket (RTR) propulsion system. Modeling of the transient heating and steady state temperatures of the system is presented. Simulation results for propellant blow down during impulsive operation are also presented. The results from this study validate the feasibility of a practical thermally capacitive RTR propulsion system.

  3. The 3D thermal-hydraulic numerical simulation for the fuel zone outlet of China experimental fast reactor

    International Nuclear Information System (INIS)

    Xue Xiuli; Yang Hongyi; Yang Fuchang

    2008-01-01

    Detailed 3D thermal-hydraulic numerical analyses to the fuel zone outlet are actualized with the STAR-CD CFD code. The performance of sodium mixing is studied and detailed velocity and temperature distribution are obtained in this region which will offer foundations and references to study the rationality of temperature monitoring-spot arrangement and to assess the effect of temperature fluctuations to control rod guide tubes in this region, and so on. (authors)

  4. Thermal-hydraulic analysis of Ignalina NPP compartments response to group distribution header rupture using RALOC4 code

    International Nuclear Information System (INIS)

    Urbonavicius, E.

    2000-01-01

    The Accident Localisation System (ALS) of Ignalina NPP is a containment of pressure suppression type designed to protect the environment from the dangerous impact of the radioactivity. The failure of ALS could lead to contamination of the environment and prescribed public radiation doses could be exceeded. The purpose of the presented analysis is to perform long term thermal-hydraulic analysis of compartments response to Group Distribution Header rupture and verify if design pressure values are not exceeded. (authors)

  5. Transient analysis and thermal hydraulic margins of GHARR-1 using the PARET/NAL code

    International Nuclear Information System (INIS)

    Adoo, N.A.

    2009-06-01

    The PARET code has been adapted by the IAEA for testing transient behaviour in research reactors. The PARET code provides a coupled thermal hydrodynamic and point kinetics capability with a continuous reactivity feedback and an optional voiding model that estimates the voiding produced by the subcooled boiling. The present version of the PARET/ANL 73 code provides a convenient means of assessing the various models and correlations proposed for the use in the analysis of research reactor behaviour. The Monte Carlo N-Particle code (MCNP) has been used to obtain power peaking profile for a two channel PARET/ANL model. A PARET model with the corresponding neutronics and thermal hydraulic characteristics for the miniature neutron source reactor (MNSR) has been used to simulate reactivity accidents for the Ghana Research Reactor - 1(GHARR-1) under the MNSR operation conditions of natural circulation, normal operation and reactivity insertion accidents. The simulation results via the insertion of large reactivity demonstrated the high inherent safety features of the MNSR for which the high negative reactivity feedback of moderator temperature limits power excursion and avoids consequently the escalation of clad temperature to the level of onset of sub-cooled void formation. The hot channel peaking factors for both radial and axial were found to be 1.17 and 1.44 respectively. Thermal hydraulic performance characteristics were investigated and the safety margins determined. The peak clad and coolant temperatures ranged from 59.18 0 C to 106.75 0 C and 42.95 0 C to 178.44 0 C respectively at which nucleate boiling will occur within the flow channels of the core. (au)

  6. Thermal-hydraulic Analysis of High-temperature Cover Gas Region in STELLA-2

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Youngchul; Son, Seok-Kwon; Yoon, Jung; Eoh, Jaehyuk; Jeong, Ji-Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The first phase of the program was focused on the key sodium component tests, and the second one has been concentrated on the sodium thermal-hydraulic integral effect test (STELLA-2). Based on its platform, simulation of the PGSFR transient will be made to evaluate plant dynamic behaviors as well as to demonstrate decay heat removal performance. Therefore, most design features of PGSFR have been modeled in STELLA-2 as closely as possible. The similarities of temperature and pressure between the model (STELLA-2) and the prototype (PGSFR) have been well preserved to reflect thermal-hydraulic behavior with natural convection as well as heat transfer between structure and sodium coolant inside the model reactor vessel (RV). For this reason, structural integrity of the entire test section should be confirmed as in the prototype. In particular, since the model reactor head in STELLA-2 supports key components and internal structures, its structural integrity exposed to high-temperature cover gas region should be confirmed. In order to reduce thermal radiation heat transfer from the hot sodium pool during normal operation, a dedicated insulation layer has been installed at the downward surface of the model reactor head to prevent direct heat flux from the sodium free surface at 545 .deg. C. Three-dimensional conjugate heat transfer analyses for the full-shape geometry of the upper part of the model reactor vessel in STELLA-2 have been carried out. Based on the results, steady-state temperature distributions in the cover gas region and the model reactor head itself have been obtained and the design requirement in temperature of the model reactor head has been newly proposed to be 350 .deg. C. For any elevated temperature conditions in STELLA-2, it was confirmed that the model reactor head generally satisfied the requirement. The CFD database constructed from this study will be used to optimize geometric parameters such as thicknesses and/or types of the insulator.

  7. Integrated assessment of thermal hydraulic processes in W7-X fusion experimental facility

    Energy Technology Data Exchange (ETDEWEB)

    Kaliatka, T., E-mail: tadas.kaliatka@lei.lt; Uspuras, E.; Kaliatka, A.

    2017-02-15

    Highlights: • The model of Ingress of Coolant Event experiment facility was developed using the RELAP5 code. • Calculation results were compared with Ingress of Coolant Event experiment data. • Using gained experience, the numerical model of Wendelstein 7-X facility was developed. • Performed analysis approved pressure increase protection system for LOCA event. - Abstract: Energy received from the nuclear fusion reaction is one of the most promising options for generating large amounts of carbon-free energy in the future. However, physical and technical problems existing in this technology are complicated. Several experimental nuclear fusion devices around the world have already been constructed, and several are under construction. However, the processes in the cooling system of the in-vessel components, vacuum vessel and pressure increase protection system of nuclear fusion devices are not widely studied. The largest amount of radioactive materials is concentrated in the vacuum vessel of the fusion device. Vacuum vessel is designed for the vacuum conditions inside the vessel. Rupture of the in-vessel components of the cooling system pipe may lead to a sharp pressure increase and possible damage of the vacuum vessel. To prevent the overpressure, the pressure increase protection system should be designed and implemented. Therefore, systematic and detailed experimental and numerical studies, regarding the thermal-hydraulic processes in cooling system, vacuum vessel and pressure increase protection system, are important and relevant. In this article, the numerical investigation of thermal-hydraulic processes in cooling systems of in-vessel components, vacuum vessels and pressure increase protection system of fusion devices is presented. Using the experience gained from the modelling of “Ingress of Coolant Event” experimental facilities, the numerical model of Wendelstein 7-X (W7-X) experimental fusion device was developed. The integrated analysis of the

  8. Test results of the new NSSS thermal-hydraulics program of the KNPEC-2 simulator

    International Nuclear Information System (INIS)

    Jeong, J. Z.; Kim, K. D.; Lee, M. S.; Hong, J. H.; Lee, Y. K.; Seo, J. S.; Kweon, K. J.; Lee, S. W.

    2001-01-01

    As a part of the KNPEC-2 Simulator Upgrade Project, KEPRI and KAERI have developed a new NSSS thermal-hydraulics program, which is based on the best-estimate system code, RETRAN. The RETRAN code was originally developed for realistic simulation of thermal-hydraulic transient in power plant systems. The capability of 'real-time simulation' and robustness' should be first developed before being implemented in full-scope simulators. For this purpose, we have modified the RETRAN code by (i) eliminating the correlations' discontinuities between flow regime maps, (ii) simplifying physical correlations, (iii) correcting errors in the original program, and (iv) others. This paper briefly presents the test results fo the new NSSS thermal-hydraulics program

  9. Thermal hydraulic considerations in liquid-metal-cooled components of tokamak fusion reactors

    International Nuclear Information System (INIS)

    Picologlou, B.F.; Reed, C.B.; Hua, T.Q.

    1989-01-01

    The basic considerations of MHD thermal hydraulics for liquid-metal-cooled blankets and first walls of tokamak fusion reactors are discussed. The liquid-metal MHD program of Argonne National Laboratory (ANL) dedicated to analytical and experimental investigations of reactor relevant MHD flows and development of relevant thermal hydraulic design tools is presented. The status of the experimental program and examples of local velocity measurements are given. An account of the MHD codes developed to date at ANL is also presented as is an example of a 3-D thermal hydraulic analysis carried out with such codes. Finally, near term plans for experimental investigations and code development are outlined. 20 refs., 8 figs., 1 tab

  10. Thermal-hydraulic analysis of PWR small assembly for irradiation test of CARR

    International Nuclear Information System (INIS)

    Yin Hao; Zou Yao; Liu Xingmin

    2015-01-01

    The thermal-hydraulic behaviors of the PWR 4 × 4 small assembly tested in the high temperature and high pressure loop of China Advanced Research Reactor were analyzed. The CFD method was used to carry out 3D simulation of the model, thus detailed thermal-hydraulic parameters were obtained. Firstly, the simplified model was simulated to give the 3D temperature and velocity distributions and analyze the heat transfer process. Then the whole scale small assembly model was simulated and the simulation results were compared with those of simplified rod bundle model. Its flow behavior was studied and flow mixing characteristics of the grids were analyzed, and the mixing factor of the grid was calculated and can be used for further thermal-hydraulic study. It is shown that the highest temperature of the fuel rod meets the design limit and the mixing effect of the grid is obvious. (authors)

  11. Modeling thermal stress propagation during hydraulic stimulation of geothermal wells

    Science.gov (United States)

    Jansen, Gunnar; Miller, Stephen A.

    2017-04-01

    A large fraction of the world's water and energy resources are located in naturally fractured reservoirs within the earth's crust. Depending on the lithology and tectonic history of a formation, fracture networks can range from dense and homogeneous highly fractured networks to single large scale fractures dominating the flow behavior. Understanding the dynamics of such reservoirs in terms of flow and transport is crucial to successful application of engineered geothermal systems (also known as enhanced geothermal systems or EGS) for geothermal energy production in the future. Fractured reservoirs are considered to consist of two distinct separate media, namely the fracture and matrix space respectively. Fractures are generally thin, highly conductive containing only small amounts of fluid, whereas the matrix rock provides high fluid storage but typically has much smaller permeability. Simulation of flow and transport through fractured porous media is challenging due to the high permeability contrast between the fractures and the surrounding rock matrix. However, accurate and efficient simulation of flow through a fracture network is crucial in order to understand, optimize and engineer reservoirs. It has been a research topic for several decades and is still under active research. Accurate fluid flow simulations through field-scale fractured reservoirs are still limited by the power of current computer processing units (CPU). We present an efficient implementation of the embedded discrete fracture model, which is a promising new technique in modeling the behavior of enhanced geothermal systems. An efficient coupling strategy is determined for numerical performance of the model. We provide new insight into the coupled modeling of fluid flow, heat transport of engineered geothermal reservoirs with focus on the thermal stress changes during the stimulation process. We further investigate the interplay of thermal and poro-elastic stress changes in the reservoir

  12. Small scale thermal-hydraulic experiment for stable operation of a pius-type reactor

    International Nuclear Information System (INIS)

    Tasaka, K.; Tamaki, M.; Imai, S.; Irianto, I.D.; Tsuji, Y.; Kukita, Y.

    1994-01-01

    Thermal-hydraulic experiments using a small-scale atmospheric pressure test loop have been performed for the Process Inherent Ultimate Safety (PIUS)-type reactor to develop the new pump speed feedback control system. Three feedback control systems based on the measurement of flow rate, differential pressure, and fluid temperature distribution in the lower density lock have been proposed and confirmed by a series of experiments. Each of the feedback control systems had been verified in the simulation experiment such as a start-up simulation test. The automatic pump speed control based on the fluid temperature at the lower density lock was quite effective to maintain the stratified interface between primary water and borated pool water for stable operation of the reactor. (author)

  13. PETER loop. Multifunctional test facility for thermal hydraulic investigations of PWR fuel elements

    International Nuclear Information System (INIS)

    Ganzmann, I.; Hille, D.; Staude, U.

    2009-01-01

    The reliable fuel element behavior during the complete fuel cycle is one of the fundamental prerequisites of a safe and efficient nuclear power plant operation. The fuel element behavior with respect to pressure drop and vibration impact cannot be simulated by means of fluid-structure interaction codes. Therefore it is necessary to perform tests using fuel element mock-ups (1:1). AREVA NP has constructed the test facility PETER (PWR fuel element tests in Erlangen) loop. The modular construction allows maximum flexibility for any type of fuel elements. Modern measuring instrumentation for flow, pressure and vibration characterization allows the analysis of cause and consequences of thermal hydraulic phenomena. PETER loop is the standard test facility for the qualification of dynamic fuel element behavior in flowing fluid and is used for failure mode analysis.

  14. Advances in thermal-hydraulic studies of a transmutation advanced device for sustainable energy applications

    International Nuclear Information System (INIS)

    Fajardo, Laura Garcia; Castells, Facundo Alberto Escriva; Lira, Carlos Brayner de Olivera

    2013-01-01

    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 trans- mutation 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, the heat transfer from the fuel to the coolant was analyzed for three core states during normal operation. The heat transfer inside the spherical fuel elements was also studied. Three critical fuel elements groups were defined regarding their position inside the core. Results were compared with a realistic CFD model of the critical fuel elements groups. During the steady state, no critical elements reached the limit temperature of this type of fuel. (author)

  15. Uncertainty propagation applied to multi-scale thermal-hydraulics coupled codes. A step towards validation

    Energy Technology Data Exchange (ETDEWEB)

    Geffray, Clotaire Clement

    2017-03-20

    The work presented here constitutes an important step towards the validation of the use of coupled system thermal-hydraulics and computational fluid dynamics codes for the simulation of complex flows in liquid metal cooled pool-type facilities. First, a set of methods suited for uncertainty and sensitivity analysis and validation activities with regards to the specific constraints of the work with coupled and expensive-to-run codes is proposed. Then, these methods are applied to the ATHLET - ANSYS CFX model of the TALL-3D facility. Several transients performed at this latter facility are investigated. The results are presented, discussed and compared to the experimental data. Finally, assessments of the validity of the selected methods and of the quality of the model are offered.

  16. Implementation of CFD module in the KORSAR thermal-hydraulic system code

    Energy Technology Data Exchange (ETDEWEB)

    Yudov, Yury V.; Danilov, Ilia G.; Chepilko, Stepan S. [Alexandrov Research Inst. of Technology (NITI), Sosnovy Bor (Russian Federation)

    2015-09-15

    The Russian KORSAR/GP (hereinafter KORSAR) computer code was developed by a joint team from Alexandrov NITI and OKB ''Gidropress'' for VVER safety analysis and certified by the Rostechnadzor of Russia in 2009. The code functionality is based on a 1D two-fluid model for calculation of two-phase flows. A 3D CFD module in the KORSAR computer code is being developed by Alexandrov NITI for representing 3D effects in the downcomer and lower plenum during asymmetrical loop operation. The CFD module uses Cartesian grid method with cut cell approach. The paper presents a numerical algorithm for coupling 1D and 3D thermal- hydraulic modules in the KORSAR code. The combined pressure field is calculated by the multigrid method. The performance efficiency of the algorithm for coupling 1D and 3D modules was demonstrated by solving the benchmark problem of mixing cold and hot flows in a T-junction.

  17. Relationship of core exit-temperature noise to thermal-hydraulic conditions in PWRs

    International Nuclear Information System (INIS)

    Sweeney, F.J.; Upadhyaya, B.R.

    1983-01-01

    Core exit thermocouple temperature noise and neutron detector noise measurements were performed at the Loss of Fluid Test Facility (LOFT) reactor and a Westinghouse, 1148 MW(e) PWR to relate temperature noise to core thermal-hydraulic conditions. The noise analysis results show that the RMS of the temperature noise increases linearly with increasing core δT at LOFT and the commercial PWR. Out-of-core test loop temperature noise has shown similar behavior. The phase angle between core exit temperature noise and in-core or ex-core neutron noise is directly related to the core coolant flow velocity. However, if the thermocouple response time is slow, compared to the coolant transit time between the sensors, velocities inferred from the phase angle are lower than measured coolant flow velocities

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

    International Nuclear Information System (INIS)

    Nishimura, M.; Ohshima, H.; Kamide, H.; Yamaguchi, K.; Yamaguchi, A.

    2000-01-01

    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)

  19. Development of a 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)

    1999-04-01

    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 a second step of the whole project, and focus to the implementation of CANDU models based on the previous study. FORTRAN 90 language have been used for the development of RELAP5.MOD3/CANDU PC version. For the convenience of the previous Workstation users, the FOTRAN 77 version has been coded also and implanted into the original RELAP5 source file. The verification of model implementation has been performed through the simple verification calculations using the CANDU version. 6 refs., 15 figs., 7 tabs. (Author)

  20. Thermal-hydraulic of partially blocked fuel subassembly with porous media

    International Nuclear Information System (INIS)

    Nagata, Takemitsu; Ohshima, Hiroyuki

    2000-10-01

    The analysis code for investigations of local subassembly phenomena, which has been recognized as an issue of local subassembly accidents, has been required and developed at JNC. It is desirable for the analysis code to be applicable to various blockage conditions and random position of the blockage formation and to evaluate conservatively on the safety assessment with high accuracy. In this study, for the purpose of verifying the application and issues of the subchannel analysis code ASFRE-IV which evaluates thermal hydraulic phenomena in the porous blockage regions, the ASFRE-IV validation analysis was carried out on the basis of the data of an experiment investigation on a local porous blockage in a fuel subassembly performed by Reactor Engineering Groop, O-arai Engineering Center, JNC. Calculational results indicated that ASFRE-IV could reproduce the coolant temperature profile in a fuel subassembly and the peak temperature in the local subchannel conservatively. (author)

  1. Simulation of Thermal-hydraulic Process in Reactor of HTR-PM

    International Nuclear Information System (INIS)

    Zhou Kefeng; Zhou Yangping; Sui Zhe; Ma Yuanle

    2014-01-01

    This paper provides the physical process in the reactor of High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) and introduces the standard operation conditions. The FORTRAN code developed for the thermal hydraulic module of Full-Scale Simulator (FSS) of HTR-PM is used to simulate two typical operation transients including cold startup process and cold shutdown process. And the results were compared to the safety analysis code, namely TINTE. The good agreement indicates that the code is applicable for simulating the thermal-hydraulic process in reactor of HTR-PM. And for long time transient process, the code shows good stability and convergence. (author)

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

    International Nuclear Information System (INIS)

    Diamond, D.J.

    1982-01-01

    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

  3. 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.

  4. CHF predictor derived from a 3D thermal-hydraulic code and an advanced statistical method

    International Nuclear Information System (INIS)

    Banner, D.; Aubry, S.

    2004-01-01

    A rod bundle CHF predictor has been determined by using a 3D code (THYC) to compute local thermal-hydraulic conditions at the boiling crisis location. These local parameters have been correlated to the critical heat flux by using an advanced statistical method based on spline functions. The main characteristics of the predictor are presented in conjunction with a detailed analysis of predictions (P/M ratio) in order to prove that the usual safety methodology can be applied with such a predictor. A thermal-hydraulic design criterion is obtained (1.13) and the predictor is compared with the WRB-1 correlation. (author)

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

    International Nuclear Information System (INIS)

    Borges, R.C.

    1980-10-01

    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) [pt

  6. Perturbative methods applied for sensitive coefficients calculations in thermal-hydraulic systems

    International Nuclear Information System (INIS)

    Andrade Lima, F.R. de

    1993-01-01

    The differential formalism and the Generalized Perturbation Theory (GPT) are applied to sensitivity analysis of thermal-hydraulics problems related to pressurized water reactor cores. The equations describing the thermal-hydraulic behavior of these reactors cores, used in COBRA-IV-I code, are conveniently written. The importance function related to the response of interest and the sensitivity coefficient of this response with respect to various selected parameters are obtained by using Differential and Generalized Perturbation Theory. The comparison among the results obtained with the application of these perturbative methods and those obtained directly with the model developed in COBRA-IV-I code shows a very good agreement. (author)

  7. BWR transient analysis using neutronic / thermal hydraulic coupled codes including uncertainty quantification

    International Nuclear Information System (INIS)

    Hartmann, C.; Sanchez, V.; Tietsch, W.; Stieglitz, R.

    2012-01-01

    The KIT is involved in the development and qualification of best estimate methodologies for BWR transient analysis in cooperation with industrial partners. The goal is to establish the most advanced thermal hydraulic system codes coupled with 3D reactor dynamic codes to be able to perform a more realistic evaluation of the BWR behavior under accidental conditions. For this purpose a computational chain based on the lattice code (SCALE6/GenPMAXS), the coupled neutronic/thermal hydraulic code (TRACE/PARCS) as well as a Monte Carlo based uncertainty and sensitivity package (SUSA) has been established and applied to different kind of transients of a Boiling Water Reactor (BWR). This paper will describe the multidimensional models of the plant elaborated for TRACE and PARCS to perform the investigations mentioned before. For the uncertainty quantification of the coupled code TRACE/PARCS and specifically to take into account the influence of the kinetics parameters in such studies, the PARCS code has been extended to facilitate the change of model parameters in such a way that the SUSA package can be used in connection with TRACE/PARCS for the U and S studies. This approach will be presented in detail. The results obtained for a rod drop transient with TRACE/PARCS using the SUSA-methodology showed clearly the importance of some kinetic parameters on the transient progression demonstrating that the coupling of a best-estimate coupled codes with uncertainty and sensitivity tools is very promising and of great importance for the safety assessment of nuclear reactors. (authors)

  8. Development of CFD software for the simulation of thermal hydraulics in advanced nuclear reactors. Final report

    International Nuclear Information System (INIS)

    Bachar, Abdelaziz; Haslinger, Wolfgang; Scheuerer, Georg; Theodoridis, Georgios

    2015-01-01

    The objectives of the project were: Improvement of the simulation accuracy for nuclear reactor thermo-hydraulics by coupling system codes with three-dimensional CFD software; Extension of CFD software to predict thermo-hydraulics in advanced reactor concepts; Validation of the CFD software by simulation different UPTF TRAM-C test cases and development of best practice guidelines. The CFD module was based on the ANSYS CFD software and the system code ATHLET of GRS. All three objectives were met: The coupled ATHLET-ANSYS CFD software is in use at GRS and TU Muenchen. Besides the test cases described in the report, it has been used for other applications, for instance the TALL-3D experiment of KTH Stockholm. The CFD software was extended with material properties for liquid metals, and validated using existing data. Several new concepts were tested when applying the CFD software to the UPTF test cases: Simulations with Conjugate Heat Transfer (CHT) were performed for the first time. This led to better agreement between predictions and data and reduced uncertainties when applying temperature boundary conditions. The meshes for the CHT simulation were also used for a coupled fluid-structure-thermal analysis which was another novelty. The results of the multi-physics analysis showed plausible results for the mechanical and thermal stresses. The workflow developed as part of the current project can be directly used for industrial nuclear reactor simulations. Finally, simulations for two-phase flows with and without interfacial mass transfer were performed. These showed good agreement with data. However, a persisting problem for the simulation of multi-phase flows are the long simulation times which make use for industrial applications difficult.

  9. Study on thermal-hydraulic phenomena in porous media. Semiannual report 1997. Nov. to 1998. Mar

    International Nuclear Information System (INIS)

    Matui, Goichi; Monji, Hideaki; Sakakibara, Jun; Tanaka, Masa-aki; Kobayashi, Jun; Kamide, Hideki

    1998-03-01

    The objective of this study is to clarify the thermal-hydraulic phenomena in porous media and to develop the analytical method to predict the thermal-hydraulic field, deciding the maximum temperature on the fuel pin surface. FY 1998 is the first year of the 3 years plan. In this year period, based on the correspondence of Reynolds number between experimental facility and FBR, the design and construction of the test rig and experimental parameter examination were performed. The test section has rectangle two sub-channel geometry and is twenty times large-scale model. In this study, we use the Particle Image Velocimetry (PIV) analysis method to visualize the flow field in the porous media. The Pyrex grass spheres were used to construct the porous blockage. The refraction-rate matching between obstacle and fluid is important to measure the velocity field with the optical analysis method. As the working fluid, NaI solution was used. When the concentration of NaI is 56.4wt% in the solution, the refraction-rate is correspond to that of the Pyrex grass. The simple test loop was constructed and the experiment was performed to measure the velocity field with Laser Doppler Anemometer and PIV. The purpose of this experiment using the simple test loop is to develop the experimental method of flow visualization in the porous media used NaI solution as working fluid. As the result of experiment with the simple test loop, the vector field in the porous media was obtained and it is shown that the flow pattern with PIV analysis is qualitatively correct. This visualization method using the NaI solution is applicable to measure the flow field in the porous media. (J.P.N.)

  10. Performance and efficiency of a hydraulic hybrid powertrain

    Energy Technology Data Exchange (ETDEWEB)

    Karbaschian, Mohammad Ali [Duisburg-Essen Univ. (Germany). Faculty of Engineering

    2012-11-01

    Hydraulic hybrid powertrains are considered to be a promising technology to save energy and reduce emission in specific automotive fields because of their high power density, components lifetime, and long lasting experience in industries compared to electric hybrid powertrains. Within the first part of the paper, a very brief literature survey on hydraulic hybrid vehicle systems (HHVS) and the related dynamical behaviour is given. No specific activities to improve the efficiency of these systems were detected. Related literature with respect to optimization mainly deals with the management of the system's energy flows trying to control the engine operation point and the high pressure in the system. In the second part, a small simulation study is presented. Therefore, hybrid systems are generally assumed as a Multi-Input-Multi-Output (MIMO) system. The effect of key variables (i.e. accumulator size and pressure, pump/motor displacement and efficiency, valve dynamics) on the system is discussed. The results show that the volume displacement of pump and motor, the performance of the engine, and the state of charge of the accumulator are the most important parameters to specify the efficiency and performance of the hydraulic hybrid powertrain. Additionally, a hybrid hydraulic powertrain with an adjustable state of charge accumulator is compared with one whose state of charge is constant. The result shows the improvement of braking performance and fuel savings. The goal is to optimize the parameters of the system based on the simultaneous consideration of the three (or more) variables for a given load profile with respect to given objectives. (orig.)

  11. Numerical solution of conservation equations in the transient model for the system thermal - hydraulics in the Korsar computer code

    International Nuclear Information System (INIS)

    Yudov, Y.V.

    2001-01-01

    The functional part of the KORSAR computer code is based on the computational unit for the reactor system thermal-hydraulics and other thermal power systems with water cooling. The two-phase flow dynamics of the thermal-hydraulic network is modelled by KORSAR in one-dimensional two-fluid (non-equilibrium and nonhomogeneous) approximation with the same pressure of both phases. Each phase is characterized by parameters averaged over the channel sections, and described by the conservation equations for mass, energy and momentum. The KORSAR computer code relies upon a novel approach to mathematical modelling of two-phase dispersed-annular flows. This approach allows a two-fluid model to differentiate the effects of the liquid film and droplets in the gas core on the flow characteristics. A semi-implicit numerical scheme has been chosen for deriving discrete analogs the conservation equations in KORSAR. In the semi-implicit numerical scheme, solution of finite-difference equations is reduced to the problem of determining the pressure field at a new time level. For the one-channel case, the pressure field is found from the solution of a system of linear algebraic equations by using the tri-diagonal matrix method. In the branched network calculation, the matrix of coefficients in the equations describing the pressure field is no longer tri-diagonal but has a sparseness structure. In this case, the system of linear equations for the pressure field can be solved with any of the known classical methods. Such an approach is implemented in the existing best-estimate thermal-hydraulic computer codes (TRAC, RELAP5, etc.) For the KORSAR computer code, we have developed a new non-iterative method for calculating the pressure field in the network of any topology. This method is based on the tri-diagonal matrix method and performs well when solving the thermal-hydraulic network problems. (author)

  12. TITAN: an advanced three-dimensional neutronics/thermal-hydraulics code for light water reactor safety analysis

    International Nuclear Information System (INIS)

    Griggs, D.P.; Kazimi, M.S.; Henry, A.F.

    1982-01-01

    The initial development of TITAN, a three-dimensional coupled neutronics/thermal-hydraulics code for LWR safety analysis, has been completed. The transient neutronics code QUANDRY has been joined to the two-fluid thermal-hydraulics code THERMIT with the appropriate feedback mechanisms modeled. A detailed steady-state and transient coupling scheme based on the tandem technique was implemented in accordance with the important structural and operational characteristics of QUANDRY and THERMIT. A two channel sample problem formed the basis for steady-state and transient analyses performed with TITAN. TITAN steady-state results were compared with those obtained with MEKIN and showed good agreement. Null transients, simulated turbine trip transients, and a rod withdrawal transient were analyzed with TITAN and reasonable results were obtained

  13. Thermal Hydraulic Design and Analysis of a Water-Cooled Ceramic Breeder Blanket with Superheated Steam for CFETR

    Science.gov (United States)

    Cheng, Xiaoman; Ma, Xuebin; Jiang, Kecheng; Chen, Lei; Huang, Kai; Liu, Songlin

    2015-09-01

    The water-cooled ceramic breeder blanket (WCCB) is one of the blanket candidates for China fusion engineering test reactor (CFETR). In order to improve power generation efficiency and tritium breeding ratio, WCCB with superheated steam is under development. The thermal-hydraulic design is the key to achieve the purpose of safe heat removal and efficient power generation under normal and partial loading operation conditions. In this paper, the coolant flow scheme was designed and one self-developed analytical program was developed, based on a theoretical heat transfer model and empirical correlations. Employing this program, the design and analysis of related thermal-hydraulic parameters were performed under different fusion power conditions. The results indicated that the superheated steam water-cooled blanket is feasible. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy of China (Nos. 2013GB108004, 2014GB122000 and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  14. Overview of the use of ATHENA for thermal-hydraulic analysis of systems with lead-bismuth coolant

    International Nuclear Information System (INIS)

    Davis, C.B.; Shieh, A. S.

    2000-01-01

    The INEEL and MIT are investigating the suitability of lead-bismuth cooled fast reactor for producing low-cost electricity as well as for actinide burning. This paper is concerned with the general area of thermal-hydraulics of lead-bismuth cooled reactors. The ATHENA code is being used in the thermal-hydraulic design and analysis of lead-bismuth cooled reactors. The ATHENA code was reviewed to determine its applicability for simulating lead-bismuth cooled reactors. Two modifications were made to the code as a result of this review. Specifically, a correlation to represent heat transfer from rod bundles to a liquid metal and a void correlation based on data taken in a mixture of lead-bismuth and steam were added the code. The paper also summarizes the analytical work that is being performed with the code and plans for future analytical work

  15. Validation of the RALOC-mod.4 thermal-hydraulics code on evaporation transients in the Phebus containment

    International Nuclear Information System (INIS)

    Spitz, P.B.; Lemoine, F.; Tirini, S.

    1997-01-01

    IPSN (Nuclear Protection and Safety Institute) and GRS (Gesellschaft fur Anlagen und Reaktorsicherheit Schwertnergasse 1) are developing the ESCADRE-ASTEC systems of codes devoted to the prediction of the behaviour of water-cooled reactors during a severe accident. The RALOC-mod 4 code belongs to this system and is specifically devoted to containment thermal-hydraulics studies. IPSN has designed a Thermal Hydraulic Containment Test Program in support to the Phebus Fission Product Test Program/2/. Evaporation tests have been recently performed in the Phebus containment test facility. The objective of this work is to assess against these tests the capability of the RALOC -mod 4 code to capture the phenomena observed in these experiments and more particularly the evaporation heat transfer and wall heat transfers. (DM)

  16. Overview of the Use of ATHENA for Thermal-Hydraulic Analysis of Systems with Lead-Bismuth Coolant

    Energy Technology Data Exchange (ETDEWEB)

    Davis, Cliff Bybee; Shieh, Arthur Shan Luk

    2000-04-01

    The INEEL and MIT are investigating the suitability of lead-bismuth cooled fast reactor for producing low-cost electricity as well as for actinide burning. This paper is concerned with the general area of thermal-hydraulics of lead-bismuth cooled reactors. The ATHENA code is being used in the thermal-hydraulic design and analysis of lead-bismuth cooled reactors. The ATHENA code was reviewed to determine its applicability for simulating lead-bismuth cooled reactors. Two modifications were made to the code as a result of this review. Specifically, a correlation to represent heat transfer from rod bundles to a liquid metal and a void correlation based on data taken in a mixture of lead-bismuth and steam were added the code. The paper also summarizes the analytical work that is being performed with the code and plans for future analytical work.

  17. Preliminary thermal-hydraulic and safety analysis of China DFLL-TBM system

    Energy Technology Data Exchange (ETDEWEB)

    Li, Wei [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Qiu, Suizheng; Su, Guanghui; Jiao, Hong [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Bai, Yunqing; Chen, Hongli [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Wu, Yican, E-mail: yican.Wu@Fds.Org.Cn [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2013-06-15

    Highlights: • Thermal-hydraulic and safety analysis on DFLL-TBM system is performed. • The TBM FW maximum temperature is 541 °C under steady state condition. • The TBM FW maximum temperature does not exceed the melt point of CLAM steel 1500 °C. • Neither the VV pressurization nor vault pressure build-up goes beyond 0.2 MPa. -- Abstract: China has proposed the dual-functional lithium-lead (DFLL) tritium breeding blanket concept for testing in ITER as a test blanket module (TBM), to demonstrate the technologies of tritium self-sufficiency, high-grade heat extraction and efficient electricity production which are needed for DEMO and fusion power plant. Safety assessment of the TBM and its auxiliary system should be conducted to deal with ITER safety issues directly caused by the TBM system failure during the design process. In this work, three potential initial events (PIEs) – in-vessel loss of helium (He) coolant and ex-vessel loss of He coolant and loss of flow without scram (LOFWS) – were analyzed for the TBM system with a modified version of the RELAP5/MOD3 code containing liquid lithium-lead eutectic (LiPb). The code also comprised an empirical expression for MHD pressure drop relevant to three-dimensional (3D) effect, the Lubarsky–Kaufman convective heat transfer correlation for LiPb flow and the Gnielinski convective heat transfer correlation for He flow. Since both LiPb and He serve as TBM coolants, the LiPb and He ancillary cooling systems were modeled to investigate the thermal-hydraulic characteristic of the TBM system and its influence on ITER safety under those accident conditions. The TBM components and the coolants flow within the TBM were simulated with one-dimensional heat structures and their associated hydrodynamic components. ITER enclosures including vacuum vessel (VV), port cell and TCWS vault were also covered in the model for accident analyses. Through this best estimate approach, the calculation indicated that the current

  18. Modelling and thermal hydraulic analysis of the Angra-2 nuclear reactor using RELAP5-3D code

    International Nuclear Information System (INIS)

    González Mantecón, Javier

    2015-01-01

    The evaluation of Nuclear Power Plants (NPPs) performance during steady-state and accident conditions has been one of the main research subjects in the nuclear field. In order to simulate the behavior of water-cooled reactors, several complex thermal-hydraulic codes systems have been developed. Particularly, the RELAP5 code, developed by the Idaho National Laboratory, is a best-estimate thermal-hydraulic analysis tool and one of the most used in nuclear industry. The RELAP5-3D 3.0.0 code was used to develop a detailed model of Angra 2 nuclear reactor using reference data from the Final Safety Analysis Report. Angra 2 is the second Brazilian NPP, which began commercial operation in 2001. The plant is equipped with a Pressurized Water Reactor (PWR) type with 3771.0 MWt. Simulations of the reactor behavior during normal operation conditions and postulated accident conditions were performed. Results achieved in the reactor steady-state simulation were compared with nominal parameters of the NPP. These results proved to be in good agreement, with relative errors less than 1%. In the transient simulation, the obtained results were coherent and satisfactory. This study demonstrates that the RELAP5-3D model is capable to reproduce the thermal-hydraulic behavior of the Angra-2 PWR during diverse operation conditions and it can contribute for the process of the plant safety analysis. (author)

  19. Thermal-hydraulic Experiments for Advanced Physical Model Development

    International Nuclear Information System (INIS)

    Song, Chulhwa

    2012-04-01

    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

  20. Basic data generation and pressure loss coefficient evaluation for HANARO core thermal-hydraulic analyses

    International Nuclear Information System (INIS)

    Chae, Hee Taek; Lee, Kye Hong

    1999-06-01

    MATRA-h, a HANARO subchannel analysis computer code, is used to evaluate thermal margin of the HANARO fuel. It's capability includes the assessments of CHF, ONB margin, and fuel temperature. In this report, basic input data and core design parameters required to perform the subchannel analysis with MATRA-h code are collected. These data include the subchannel geometric data, thermal-hydraulic correlations, empirical constants and material properties. The friction and form loss coefficients of the fuel assemblies were determined based on the results of the pressure drop test. At the same time, different form loss coefficients at the end plates and spacers are evaluated for various subchannels. The adequate correlations are applied to the evaluation of the form loss coefficients for various subchannels, which are corrected by measured values in order to have a same pressure drop at each flow channel. These basic input data and design parameters described in this report will be applied usefully to evaluate the thermal margin of the HANARO fuel. (author). 11 refs., 13 tabs., 11 figs

  1. Transient thermal hydraulic modeling and analysis of ITER divertor plate system

    International Nuclear Information System (INIS)

    El-Morshedy, Salah El-Din; Hassanein, Ahmed

    2009-01-01

    A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m 2 plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.

  2. Transient thermal hydraulic modeling and analysis of ITER divertor plate system

    Energy Technology Data Exchange (ETDEWEB)

    El-Morshedy, Salah El-Din [Argonne National Laboratory, Argonne, IL (United States); Atomic Energy Authority, Cairo (Egypt)], E-mail: selmorshedy@etrr2-aea.org.eg; Hassanein, Ahmed [Purdue University, West Lafayette, IN (United States)], E-mail: hassanein@purdue.edu

    2009-12-15

    A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m{sup 2} plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.

  3. Thermal hydraulic and power cycle analysis of liquid lithium blanket designs

    International Nuclear Information System (INIS)

    Misra, B.; Stevens, H.C.; Maroni, V.A.

    1977-01-01

    Thermal hydraulic and power cycle analyses were performed for the first-wall and blanket systems of tokamak-type fusion reactors under a typical set of design and operating conditions. The analytical results for lithium-cooled blanket cells show that with stainless steel as construction material and with no divertor present, the maximum allowable neutron wall loading is approximately 2 MW/m 2 and is limited by thermal stress criteria. With vanadium alloy as construction material and no divertor present, the maximum allowable neutron wall loading is approximately 8 MW/m 2 and is limited by an interplay of constraints imposed on the maximum allowable structural temperature and the minimum allowable coolant inlet temperature. With a divertor these wall loadings can be increased by from 40 to 90 percent. The cost of the vanadium system is found to be competitive with the stainless steel system because of the higher allowable structural temperatures and concomitant higher thermal efficiencies afforded by the vanadium alloys

  4. Development of whole core thermal-hydraulic analysis program ACT. 4. Incorporation of three-dimensional upper plenum model

    International Nuclear Information System (INIS)

    Ohshima, Hiroyuki

    2003-03-01

    The thermal-hydraulic analysis computer program ACT is under development for the evaluation of detailed flow and temperature fields in a core region of fast breeder reactors under various operation conditions. The purpose of this program development is to contribute not only to clarifying thermal hydraulic characteristics that cannot be revealed by experiments due to measurement difficulty but also to performing rational safety design and assessment. This report describes the incorporation of a three-dimensional upper plenum model to ACT and its verification study as part of the program development. To treat the influence of three-dimensional thermal-hydraulic behavior in a upper plenum on the in-core temperature field, the multi-dimensional general purpose thermal-hydraulic analysis program AQUA, which was developed and validated at JNC, was applied as the base of the upper plenum analysis module of ACT. AQUA enables to model the upper plenum configuration including immersed heat exchangers of the direct reactor auxiliary cooling system (DRACS). In coupling core analysis module that consists of the fuel-assembly and the inter-wrapper gap calculation parts with the upper plenum module, different types of computation mesh systems were jointed using the staggered quarter assembly mesh scheme. A coupling algorithm among core, upper plenum and heat transport system modules, which can keep mass, momentum and energy conservation, was developed and optimized in consideration of parallel computing. ACT was applied to analyzing a sodium experiment (PLANDTL-DHX) performed at JNC, which simulated the natural circulation decay heat removal under DRACS operation conditions for the program verification. From the calculation result, the validity of the improved program was confirmed. (author)

  5. Thermal hydraulic modelling of the Mo and Iridium irradiation facilities of the RA10 reactor

    International Nuclear Information System (INIS)

    Gramajo, M.; García, J.; Marcel, C.P.

    2013-01-01

    The RA-10 reactor is a multipurpose, open pool research reactor. The core consists of a rectangular array of MTR type fuel. The produced thermal power is 30 MW which is extracted by the refrigeration system via an ascendant flow through the core. The core reflector is D 2 O contained in a watertight tank. The design of the reactor includes a number of out-core facilities which are meant to be used for industrial, medical and research purposes. Among all the facilities, the most important ones are the Molybdenum and Iridium ones which we modeled in this work. During the normal operation of the reactor, the manipulation and the on-line extraction of the irradiation facilities is foreseen. Therefore the study of the head loss during the normal operation as well as during the extraction maneuvers plays a relevant role in the design and safety analysis. In this work a CFD commercial code is use dto perform the calculations needed to guarantee the design requirements.In addition, a full detailed geometric model for both, the Molybdenum and Iridium facilities,is used to perform the required simulations. The obtained results allow to evaluating the thermal-hydraulic performance of the proposed facilities designs. (author)

  6. Engineering and thermal-hydraulic design of water cooled PFC for SST-1 tokamak

    International Nuclear Information System (INIS)

    Paritosh Chaudhuri; Santra, P.; Rabi Prakash, N.; Khirwadkar, S.; Arun Prakash, A.; Ramash, G.; Dubey, S.; Chenna Reddy, D.; Saxena, Y.C.

    2005-01-01

    plate is necessary for the efficient heat transfer from the tube to the back plate. The contact at the brazed joint of the tube to the backplate/heat sink is critical for the above application. The manufactured modules need to be evaluated for the quality of brazed joint. Using an infra-red-camera, spatial and temporal evaluation of the temperature profile has been studied under various flow parameters. Design considerations included 2-D steady state and transient tile temperature distribution and resulting thermal loads in PFC during cooling, coolant parameters necessary to maintain optimum thermal-hydraulic design, erosion rate of the tiles, and tile fitting mechanism. A 2-D Finite Difference code has been developed to study of flow behavior and thermal response of PFC during cooling. The temperature distribution results for different PFC obtained by code were assessed by comparison with 2-D Finite Element (FE) method (using ANSYS). FE models have been developed to conduct the heat transfer and stress analyses of the PFC to understand its thermal and mechanical behaviors. The result of the calculation led to a good understanding of the flow behavior and the temperature distribution in the tube wall and the different parts of the PFC. Thermal response on divertors has been performed both in steady state and transient case. Stress analyses also have been performed by ANSYS to investigate the thermal stress on different PFC during cooling. In this paper an optimized thermal-hydraulic design of PFC cooling and their thermal response will be discussed in detail. (authors)

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

    International Nuclear Information System (INIS)

    Caruso, R.

    1997-01-01

    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 Design Basis Accidents, , 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

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

    International Nuclear Information System (INIS)

    Lee, Y. G.; Kim, J. W.; Yoon, S. J.; Park, G. C.

    2010-10-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-10-15

    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)

  10. 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.

  11. Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD

    Energy Technology Data Exchange (ETDEWEB)

    Trambauer, K. [GRS, Garching (Germany)

    1997-07-01

    The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonable accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.

  12. Validation of the Thermal-Hydraulic Model in the SACAP Code with the ISP Tests

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soon-Ho; Kim, Dong-Min; Park, Chang-Hwan [FNC Technology Co., Yongin (Korea, Republic of)

    2016-10-15

    In safety viewpoint, the pressure of the containment is the important parameter, of course, the local hydrogen concentration is also the parameter of the major concern because of its flammability and the risk of the detonation. In Korea, there have been an extensive efforts to develop the computer code which can analyze the severe accident behavior of the pressurized water reactor. The development has been done in a modularized manner and SACAP(Severe Accident Containment Analysis Package) code is now under final stage of development. SACAP code adopts LP(Lumped Parameter) model and is applicable to analyze the synthetic behavior of the containment during severe accident occurred by thermal-hydraulic transient, combustible gas burn, direct containment heating by high pressure melt ejection, steam explosion and molten core-concrete interaction. The analyses of a number of ISP(International Standard Problem) experiments were done as a part of the SACAP code V and V(verification and validation). In this paper, the SACAP analysis results for ISP-35 NUPEC and ISP-47 TOSQAN are presented including comparison with other existing NPP simulation codes. In this paper, we selected and analyzed ISP-35 NUPEC, ISP-47 TOSQAN in order to confirm the computational performance of SACAP code currently under development. Now the multi-node analysis for the ISP-47 is under process. As a result of simulation, SACAP predicts well the thermal-hydraulic variables such as temperature, pressure, etc. Also, we verify that SACAP code is properly equipped to analyze the gas distribution and condensation.

  13. Thermal Hydraulic Analysis of K-DEMO Single Blanket Module for Preliminary Accident Analysis using MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Sung Bo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2016-05-15

    To develop the Korean fusion commercial reactor, preliminary design concept for K-DEMO (Korean fusion demonstration reactor) has been announced by NFRI (National Fusion Research Institute). This pre-conceptual study of K-DEMO has been introduced to identify technical details of a fusion power plant for the future commercialization of fusion reactor in Korea. Before this consideration, to build the K-DEMO, accident analysis is essential. Since the Fukushima accident, which is severe accident from unexpected disaster, safety analysis of nuclear power plant has become important. The safety analysis of both fission and fusion reactors is deemed crucial in demonstrating the low radiological effect of these reactors on the environment, during severe accidents. A risk analysis of K-DEMO should be performed, as a prerequisite for the construction of a fusion reactor. In this research, thermal-hydraulic analysis of single blanket module of K-DEMO is conducted for preliminary accident analysis for K-DEMO. Further study about effect of flow distributer is conducted. The normal K-DEMO operation condition is applied to the boundary condition and simulated to verify the material temperature limit using MELCOR. MELCOR is fully integrated, relatively fast-running code developed by Sandia National Laboratories. MELCOR had been used for Light Water Reactors and fusion reactor version of MELCOR was developed for ITER accident analysis. This study shows the result of thermal-hydraulic simulation of single blanket module with MELCOR which is severe accident code for nuclear fusion safety analysis. The difference of mass flow rate for each coolant channel with or without flow distributer is presented. With flow distributer, advantage of broadening temperature gradient in the K-DEMO blanket module and increase mass flow toward first wall is obtained. This can enhance the safety of K-DEMO blanket module. Most 13 .deg. C temperature difference in blanket module is obtained.

  14. Study on thermal performance and margins of BWR fuel elements

    International Nuclear Information System (INIS)

    Stosic, Zoran

    1999-01-01

    This paper contributes to developing a methodology of predicting and analyzing thermal performance and margins of Boiling Water Reactor (BWR) fuel assemblies under conditions of reaching high quality Boiling Crisis and subsequent post-dryout thermal hydraulics causing temperature excursion of fuel cladding. Operational margins against dryout and potential for increasing fuel performance with appropriate benefits are discussed. The philosophy of modeling with its special topics are demonstrated on the HECHAN (HEated CHannel ANalyzer) model as the state-of-art for thermal-hydraulics analysis of BWR fuel assemblies in pre- and post-dryout two-phase flow regimes. The scope of further work either being or has to be performed concerning implementation of new physical aspects, including domain extension of HECHAN model applications to the Pressurized Water Reactors (PWRs), is discussed. Finally, a comprehensive overview of the literature dealing with development of the model is given. (author)

  15. 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

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

    International Nuclear Information System (INIS)

    Choi, Sun Rock; Cho, Chung Ho; Kim, Sang Ji

    2011-01-01

    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

  17. Review of thermal-hydraulic calculations for Calvert Cliffs and H.B. Robinson PTS study

    International Nuclear Information System (INIS)

    Jo, J.H.; Yuelys-Miksis, C.; Rohatgi, U.S.

    1984-01-01

    Thermal-hydraulic transient calculations performed by LANL using the TRAC-PF1 code and by INEL using the RELAP5 code for the USNRC pressurized thermal shock (PTS) study of the Calvert Cliffs and H.B. Robinson Nuclear Power Plants have been reviewed at BNL including the input decks and steady state calculations. Furthermore, six transients for each plant have been selected for the in-depth review. Simple hand calculations based on the mass and energy balances of the entire reactor system, have been performed to predict the temperature and pressure of the reactor system, and the results have been compared with those obtained by the code calculation. In general, the temperatures and pressures of the primary system calculated by the codes have been very reasonable. The secondary pressures calculated by TRAC appear to indicate that the codes have some difficulty with the condensation model and further work is needed to assess the code calculation of the U-tube steam generator pressure when the cold auxiliary feedwater is introduced to the steam generator. However, it is not expected that this uncertainty would affect the transient calculations significantly

  18. Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

    International Nuclear Information System (INIS)

    Costa, Antonella L.; Reis, Patricia Amelia L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Mesquita, Amir Z.; Soares, Humberto V.

    2010-01-01

    The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.

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

    International Nuclear Information System (INIS)

    Zhou, Jianjun; Zhang, Daling; Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei

    2015-01-01

    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. VIPRE-01: a thermal-hydraulic code for reactor cores. Volume 3: programmer's manual (Revision 2)

    International Nuclear Information System (INIS)

    Stewart, C.W.; Koontz, A.S.; Cuta, J.M.; Montgomery, S.D.

    1985-07-01

    The VIPRE thermal-hydraulic computer code for PWR and BWR core analysis has undergone a detailed design review by a committee of experts. A new version of the code, incorporating the committee's recommendations, has been submitted for NRC review and issuance of a safety evaluation report. The changes in the programmers's manual are given

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

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Jianjun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); College of Mechanical and Power Engineering, China Three Gorges University, No 8, Daxue road, Yichang, Hubei 443002 (China); Zhang, Daling, E-mail: dlzhang@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China)

    2015-02-15

    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.

  2. Thermal-hydraulic feedback model to calculate the neutronic cross-section in PWR reactions

    International Nuclear Information System (INIS)

    Santiago, Daniela Maiolino Norberto

    2011-01-01

    In neutronic codes,it is important to have a thermal-hydraulic feedback module. This module calculates the thermal-hydraulic feedback of the fuel, that feeds the neutronic cross sections. In the neutronic co de developed at PEN / COPPE / UFRJ, the fuel temperature is obtained through an empirical model. This work presents a physical model to calculate this temperature. We used the finite volume technique of discretized the equation of temperature distribution, while calculation the moderator coefficient of heat transfer, was carried out using the ASME table, and using some of their routines to our program. The model allows one to calculate an average radial temperature per node, since the thermal-hydraulic feedback must follow the conditions imposed by the neutronic code. The results were compared with to the empirical model. Our results show that for the fuel elements near periphery, the empirical model overestimates the temperature in the fuel, as compared to our model, which may indicate that the physical model is more appropriate to calculate the thermal-hydraulic feedback temperatures. The proposed model was validated by the neutronic simulator developed in the PEN / COPPE / UFRJ for analysis of PWR reactors. (author)

  3. Validation of a thermal-hydraulic system code on a simple example

    International Nuclear Information System (INIS)

    Kopecek, Vit; Zacha, Pavel

    2014-01-01

    A mathematical model of a U tube was set up and the analytical solution was calculated and used in the assessment of the numerical solutions obtained by using the RELAP5 mod3.3 and TRACE V5 thermal hydraulics codes. A good agreement between the 2 types of calculation was obtained.

  4. Trend analysis of troubles caused by thermal-hydraulic phenomena at nuclear power plants

    International Nuclear Information System (INIS)

    Komatsu, Teruo

    2010-01-01

    The Institute of Nuclear Safety System (INSS) is promoting researches to improve the safety and reliability of nuclear power plants. In the present study, our attention was focused on troubles attributed to thermal-hydraulic phenomena in particular, trend analysis were carried out to learn lessons from these troubles and to prevent their recurrence. Through our survey, we found the following two points. First, many thermal-hydraulics related troubles can be attributed to design faults, since we found some events in foreign countries took place after inadequate facility renovation. To ensure appropriate design verification, it is important to take account of state-of-the-art science and technology and at the same time to pay attention to the compatibility with the initial design concept. Second point, thermal-hydraulic related troubles are common and recurrent to nuclear power plants worldwide. Japanese utilities are planning to introduce some of overseas experiences to their plants, such as power uprate and renovations of aged facilities. It is important to learn lessons from experiences paying close attention continuously to overseas trouble events, including thermal-hydraulics related events, and to use them to improve safety and reliability of nuclear power plants. (author)

  5. Advances in thermal hydraulic and neutronic simulation for reactor analysis and safety

    International Nuclear Information System (INIS)

    Tentner, A.M.; Blomquist, R.N.; Canfield, T.R.; Ewing, T.F.; Garner, P.L.; Gelbard, E.M.; Gross, K.C.; Minkoff, M.; Valentin, R.A.

    1993-01-01

    This paper describes several large-scale computational models developed at Argonne National Laboratory for the simulation and analysis of thermal-hydraulic and neutronic events in nuclear reactors and nuclear power plants. The impact of advanced parallel computing technologies on these computational models is emphasized

  6. Thermal-hydraulic calculation and analysis for QNPP (Qinshan Nuclear Power Plant) containment

    International Nuclear Information System (INIS)

    Xie Hui; Zhou Jie; He Yingchao

    1993-01-01

    Three containment thermal-hydraulic codes CONTEMPT-LT/028, CONTEMPT-4/MOD3 and COMPARE are used to compute and analyse the Qinshan Nuclear Power Plant (QNPP) containment response under LOCA or MSLB conditions. An evaluation of the capability of containment of QNPP is given

  7. Characteristic thermal-hydraulic problems in NHRs: Overview of experimental investigations and computer codes

    Energy Technology Data Exchange (ETDEWEB)

    Falikov, A A; Vakhrushev, V V; Kuul, V S; Samoilov, O B; Tarasov, G I [OKBM, Nizhny Novgorod (Russian Federation)

    1997-09-01

    The paper briefly reviews the specific thermal-hydraulic problems for AST-type NHRs, the experimental investigations that have been carried out in the RF, and the design procedures and computer codes used for AST-500 thermohydraulic characteristics and safety validation. (author). 13 refs, 10 figs, 1 tab.

  8. Transient thermal-hydraulic/neutronic analysis in a VVER-1000 reactor core

    International Nuclear Information System (INIS)

    Seyed khalil Mousavian; Mohammad Mohsen Ertejaei; Majid Shahabfar

    2005-01-01

    Full text of publication follows: Nowadays, coupled thermal-hydraulic and three-dimensional neutronic codes in order to consider different feedback effects is state of the art subject in nuclear engineering researches. In this study, RELAP5/COBRA and WIMS/CITATION codes are implemented to investigate the VVER-1000 reactor core parameters during Large Break Loss of Coolant Accident (LB-LOCA). In a LB-LOCA, the primary side pressure, coolant density and fuel temperature strongly decrease but the cladding temperature experiences a strong peak. For this purpose, the RELAP5 Best Estimate (BE) system code is used to simulate the LB-LOCA analysis in VVER-1000 nuclear thermal-hydraulic loops. Also, the modified COBRA-IIIc software as a sub-channel analysis code is applied for modeling of VVER-1000 reactor core. Moreover, WIMS and CITATION as a cross section and 3-D neutron flux codes are coupled with thermal-hydraulic codes with the aim of consider the spatial effects through the reactor core. For this reason, suitable software is developed to link and speed up the coupled thermalhydraulic and three-dimensional neutronic calculations. This software utilizes of external coupling concept in order to integrate thermal-hydraulic and neutronic calculations. (authors)

  9. Steady-state thermal-hydraulic analysis of the pellet-bed reactor for nuclear thermal propulsion

    International Nuclear Information System (INIS)

    El-Genk, M.S.; Morley, N.J.; Yang, J.Y.

    1992-01-01

    The pellet-bed reactor (PBR) for nuclear thermal propulsion is a hydrogen-cooled, BeO-reflected, fast reactor, consisting of an annular core region filled with randomly packed, spherical fuel pellets. The fuel pellets in the PBR are self-supported, eliminating the need for internal core structure, which simplifies the core design and reduces the size and mass of the reactor. Each spherical fuel pellet is composed of hundreds of fuel microspheres embedded in a zirconium carbide (ZrC) matrix. Each fuel microsphere is composed of a UC-NbC fuel kernel surrounded by two consecutive layers of the NbC and ZrC. Gaseous hydrogen serves both as core coolant and as the propellant for the PBR rocket engine. The cold hydrogen flows axially down the inlet channel situated between the core and the external BeO reflector and radially through the orifices in the cold frit, the core, and the orifices in the hot frit. Finally, the hot hydrogen flows axially out the central channel and exits through converging-diverging nozzle. A thermal-hydraulic analysis of the PBR core was performed with an emphasis on optimizing the size and axial distribution of the orifices in the hot and cold frits to ensure that hot spots would not develop in the core during full-power operation. Also investigated was the validity of the assumptions of neglecting the axial conduction and axial cross flow in the core

  10. A Newton-based Jacobian-free approach for neutronic-Monte Carlo/thermal-hydraulic static coupled analysis

    International Nuclear Information System (INIS)

    Mylonakis, Antonios G.; Varvayanni, M.; Catsaros, N.

    2017-01-01

    Highlights: •A Newton-based Jacobian-free Monte Carlo/thermal-hydraulic coupling approach is introduced. •OpenMC is coupled with COBRA-EN with a Newton-based approach. •The introduced coupling approach is tested in numerical experiments. •The performance of the new approach is compared with the traditional “serial” coupling approach. -- Abstract: In the field of nuclear reactor analysis, multi-physics calculations that account for the bonded nature of the neutronic and thermal-hydraulic phenomena are of major importance for both reactor safety and design. So far in the context of Monte-Carlo neutronic analysis a kind of “serial” algorithm has been mainly used for coupling with thermal-hydraulics. The main motivation of this work is the interest for an algorithm that could maintain the distinct treatment of the involved fields within a tight coupling context that could be translated into higher convergence rates and more stable behaviour. This work investigates the possibility of replacing the usually used “serial” iteration with an approximate Newton algorithm. The selected algorithm, called Approximate Block Newton, is actually a version of the Jacobian-free Newton Krylov method suitably modified for coupling mono-disciplinary solvers. Within this Newton scheme the linearised system is solved with a Krylov solver in order to avoid the creation of the Jacobian matrix. A coupling algorithm between Monte-Carlo neutronics and thermal-hydraulics based on the above-mentioned methodology is developed and its performance is analysed. More specifically, OpenMC, a Monte-Carlo neutronics code and COBRA-EN, a thermal-hydraulics code for sub-channel and core analysis, are merged in a coupling scheme using the Approximate Block Newton method aiming to examine the performance of this scheme and compare with that of the “traditional” serial iterative scheme. First results show a clear improvement of the convergence especially in problems where significant

  11. Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

    International Nuclear Information System (INIS)

    Cho, Jaehyun; Park, Jin Hee; Kim, Dong-San; Lim, Ho-Gon

    2017-01-01

    Highlights: • We quantified the LOCA core damage frequency based on the best-estimated success criteria analysis. • The thermal-hydraulic analysis using MARS code has been applied to Korea Standard Nuclear Power Plants. • Five new event trees with new break size boundaries and new success criteria were developed. • The core damage frequency is 5.80E−07 (/y), which is 12% less than the conventional PSA event trees. - Abstract: A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4 in and by only a high pressure safety injection in LOCAs of 0.8–9.4 in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety

  12. Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jaehyun, E-mail: chojh@kaeri.re.kr; Park, Jin Hee; Kim, Dong-San; Lim, Ho-Gon

    2017-04-15

    Highlights: • We quantified the LOCA core damage frequency based on the best-estimated success criteria analysis. • The thermal-hydraulic analysis using MARS code has been applied to Korea Standard Nuclear Power Plants. • Five new event trees with new break size boundaries and new success criteria were developed. • The core damage frequency is 5.80E−07 (/y), which is 12% less than the conventional PSA event trees. - Abstract: A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4 in and by only a high pressure safety injection in LOCAs of 0.8–9.4 in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety

  13. Thermal-hydraulic limitations on water-cooled limiters

    International Nuclear Information System (INIS)

    Cha, Y.S.; Misra, B.

    1984-08-01

    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

  14. Three-dimensional coupled kinetics/thermal- hydraulic benchmark TRIGA experiments

    International Nuclear Information System (INIS)

    Feltus, Madeline Anne; Miller, William Scott

    2000-01-01

    This research project provides separate effects tests in order to benchmark neutron kinetics models coupled with thermal-hydraulic (T/H) models used in best-estimate codes such as the Nuclear Regulatory Commission's (NRC) RELAP and TRAC code series and industrial codes such as RETRAN. Before this research project was initiated, no adequate experimental data existed for reactivity initiated transients that could be used to assess coupled three-dimensional (3D) kinetics and 3D T/H codes which have been, or are being developed around the world. Using various Test Reactor Isotope General Atomic (TRIGA) reactor core configurations at the Penn State Breazeale Reactor (PSBR), it is possible to determine the level of neutronics modeling required to describe kinetics and T/H feedback interactions. This research demonstrates that the small compact PSBR TRIGA core does not necessarily behave as a point kinetics reactor, but that this TRIGA can provide actual test results for 3D kinetics code benchmark efforts. This research focused on developing in-reactor tests that exhibited 3D neutronics effects coupled with 3D T/H feedback. A variety of pulses were used to evaluate the level of kinetics modeling needed for prompt temperature feedback in the fuel. Ramps and square waves were used to evaluate the detail of modeling needed for the delayed T/H feedback of the coolant. A stepped ramp was performed to evaluate and verify the derived thermal constants for the specific PSBR TRIGA core loading pattern. As part of the analytical benchmark research, the STAR 3D kinetics code (, STAR: Space and time analysis of reactors, Version 5, Level 3, Users Guide, Yankee Atomic Electric Company, YEAC 1758, Bolton, MA) was used to model the transient experiments. The STAR models were coupled with the one-dimensional (1D) WIGL and LRA and 3D COBRA (, COBRA IIIC: A digital computer program for steady-state and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements, Battelle

  15. Verification Test of Hydraulic Performance for Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Jun; Kim, Jae Shin; Ryu, In Wan; Ko, Bok Seong; Song, Keun Myung [Samjin Ind. Co., Seoul (Korea, Republic of)

    2010-01-15

    According to this project, basic design for prototype pump and model pump of reactor coolant pump and test facilities has been completed. Basic design for prototype pump to establish structure, dimension and hydraulic performance has been completed and through primary flow analysis by computational fluid dynamics(CFD), flow characteristics and hydraulic performance have been established. This pump was designed with mixed flow pump having the following design requirements; specific velocity(Ns); 1080.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 3115m{sup 3}/h, total head ; 26.3m, pump speed; 1710rpm, pump efficiency; 77.0%, Impeller out-diameter; 349mm, motor output; 360kw, design pressure; 17MPaG. The features of the pump are leakage free due to no mechanical seal on the pump shaft which insures reactor's safety and law noise level and low vibration due to no cooling fan on the motor which makes eco-friendly product. Model pump size was reduced to 44% of prototype pump for the verification test for hydraulic performance of reactor coolant pump and was designed with mixed flow pump and canned motor having the following design requirements; specific speed(NS); 1060.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 539.4m{sup 3}/h, total head; 21.0m, pump speed; 3476rpm, pump efficiency; 72.9%, Impeller out-diameter; 154mm, motor output; 55kw, design pressure; 1.0MPaG. The test facilities were designed for verification test of hydraulic performance suitable for pump performance test, homologous test, NPSH test(cavitation), cost down test and pressure pulsation test of inlet and outlet ports. Test tank was designed with testing capacity enabling up to 2000m{sup 3}/h and design pressure 1.0MPaG. Auxiliary pump was designed with centrifugal pump having capacity; 1100m{sup 3}/h, total head; 42.0m, motor output; 190kw

  16. Thermal-Hydraulic Results for the Boiling Water Reactor Dry Cask Simulator.

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-09-01

    The thermal performance of commercial nuclear spent fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both aboveground and belowground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. The purpose of this investigation was to produce validation-quality data that can be used to test the validity of the modeling presently used to determine cladding temperatures in modern vertical dry casks. These cladding temperatures are critical to evaluate cladding integrity throughout the storage cycle. To produce these data sets under well-controlled boundary conditions, the dry cask simulator (DCS) was built to study the thermal-hydraulic response of fuel under a variety of heat loads, internal vessel pressures, and external configurations. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly was deployed inside of a representative storage basket and cylindrical pressure vessel that represents a vertical canister system. The symmetric

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

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki; Ohnuki, Akira; Akimoto, Hajime

    2000-03-01

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

  18. Thermal-hydraulic experiment and analysis for interim dry storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Yoo, Seung Hun

    2011-02-01

    The experimental and numerical studies of interim storages for nuclear spent fuels have been performed to investigate thermal-hydraulic characteristics of the dry storage systems and to propose new methodologies for the analysis and the design. Three separate researches have been performed in the present study: (a) Development of a scaling methodology and thermal-hydraulic experiment of a single spent fuel assembly simulating a dry storage cask: (b) Full-scope simulation of a dry storage cask by the use of Computational Fluid Dynamics (CFD) code: (c) Thermal-hydraulic design of a tunnel-type interim storage facility. In the first study, a scaling methodology has been developed to design a scaled-down canister. The scaling was performed in two steps. For the first step, the height of a spent fuel assembly was reduced from full height to half height. In order to consider the effect of height reduction on the natural convection, the scaling law of Ishii and Kataoka (1984) was employed. For the second step, the quantity of spent fuel assemblies was reduced from multiple assemblies to a single assembly. The scaling methodology was validated through the comparison with the experiment of the TN24P cask. The Peak Cladding Temperature (PCT), temperature gradients, and the axial and radial temperature distribution in the nondimensional forms were in good agreement with the experimental data. Based on the developed methodology, we have performed a single assembly experiment which was designed to simulate the full scale of the TN24P cask. The experimental data was compared with the CFD calculations. It turns out that their PCTs were less than the maximum allowable temperature for the fuel cladding and that the differences of their PCTs were agreed within 3 .deg. C, which was less than measurement uncertainty. In the second study, the full-scope simulations of the TN24P cask were performed by FLUENT. In order to investigate the sensitivity of the numerical and physical

  19. Analysis of molten salt thermal-hydraulics using computational fluid dynamics

    International Nuclear Information System (INIS)

    Yamaji, B.; Csom, G.; Aszodi, A.

    2003-01-01

    To give a good solution for the problem of high level radioactive waste partitioning and transmutation is expected to be a pro missing option. Application of this technology also could extend the possibilities of nuclear energy. Large number of liquid-fuelled reactor concepts or accelerator driven subcritical systems was proposed as transmutors. Several of these consider fluoride based molten salts as the liquid fuel and coolant medium. The thermal-hydraulic behaviour of these systems is expected to be fundamentally different than the behaviour of widely used water-cooled reactors with solid fuel. Considering large flow domains three-dimensional thermal-hydraulic analysis is the method seeming to be applicable. Since the fuel is the coolant medium as well, one can expect a strong coupling between neutronics and thermal-hydraulics too. In the present paper the application of Computational Fluid Dynamics for three-dimensional thermal-hydraulics simulations of molten salt reactor concepts is introduced. In our past and recent works several calculations were carried out to investigate the capabilities of Computational Fluid Dynamics through the analysis of different molten salt reactor concepts. Homogenous single region molten salt reactor concept is studied and optimised. Another single region reactor concept is introduced also. This concept has internal heat exchanges in the flow domain and the molten salt is circulated by natural convection. The analysis of the MSRE experiment is also a part of our work since it may form a good background from the validation point of view. In the paper the results of the Computational Fluid Dynamics calculations with these concepts are presented. In the further work our objective is to investigate the thermal-hydraulics of the multi-region molten salt reactor (Authors)

  20. Thermal and hydraulic analyses of TFTR cooling water system and magnetic field coils

    International Nuclear Information System (INIS)

    Lee, A.Y.

    1975-10-01

    The TFTR toroidal field coils, ohmic heating, hybrid and equilibrium field coils are cooled by water from the machine area cooling water system. The system has the following major equipment and capacities: flow rate of 3600 gpm; ballast tank volume of 5500 gal; pumps of 70.4 m head; chiller refrigeration rating of 3300 tons and connecting pipe of 45.7 cm I.D. The performance of the closed loop system was analyzed and found to be adequate for the thermal loads. The field coils were analyzed with detailed thermal and hydraulic models, including a simulation of the complete water cooling loop. Under the nominal operating mode of one second of toroidal field flat top time and 300 seconds of pulse cycle time, the maximum temperature for the TF coils is 53 0 C; for the OH coils 46 0 C and for the EF coils 39 0 C, which are well below the coil design limit of 120 0 C. The maximum TF coil coolant temperature is 33 0 C which is below the coolant design limit of 100 0 C. The overall pressure loss of the system is below 6.89 x 10 5 Pa (100 psi). With the given chiller refrigeration capacity, the TF coils can be operated to yield up to 4 seconds of flat top time. The TF coils can be operated on a steady state basis at up to 20% of the pulsed duty design current rating of 7.32 kA/coil

  1. Lead-Bismuth Eutectic cooled experimental Accelerator Driven System. Windowless target unit thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Bianchi, F.; Ferri, R.; Moreau, V.

    2004-01-01

    A main concern related to the peaceful use of nuclear energy is the safe management of nuclear wastes, with particular attention to long-lived fission products. An increasing attention has recently been addressed to transmutation systems (Accelerator Driven System: ADS) able to 'burn' the actinides and some of the long-lived fission products (High-Level Waste: HLW), transforming them in short or medium-lived wastes that may be easier managed and stored in the geological disposal, with the consequent easier acceptability by population. An ADS consists of a subcritical-core coupled with an accelerator by means of a target. This paper deals with the thermal-hydraulic analysis, performed with STAR-CD and RELAP5 codes for the windowless target unit of Lead-Bismuth Eutectic (LBE) cooled experimental ADS (XADS), both to assess its behaviour during operational and accident sequences and to provide input data for the thermal-mechanical analyses. It also reports a description of modifications properly implemented in the codes used for the assessment of this kind of plants. (author)

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  3. Research and development program for PWR safety at the CEA reactor thermal hydraulics laboratories

    International Nuclear Information System (INIS)

    Bernard, M.

    1995-01-01

    Since the start of the French electronuclear program, the three partners Fermate, EDF and Cea (DRN and IPSN) have devoted considerable effort to research and development for safety issues. In particular an important program on thermal hydraulics was initiated at the beginning of the seventies. It is illustrated by the development of the CATHARE thermalhydraulic safety code which includes physical models derived from a large experimental support program and the construction of the BETHSY integral facility which is aimed to assess both the CATHARE code and the physical relevance of the accident management procedures to be applied on reactors. The state of the art on this program is described with particular emphasis on the capabilities and the assessment of the last version of CATHARE and the lessons drawn from 50 BETHSY tests performed so far. The future plans for safety research cover the following strategy: - to solve the few problems identified on present computing tools and extend the assessment - to solve the few problems identified on present computing tools and extend the assessment - to perform safety studies on the basis of plant operation feedback - to contribute to treating the safety issues related to the future reactors and in particular the case of severe accidents which have to be taken into account from the design stage. The program on severe accidents is aimed to support the design studies performed by the industrial partners and to provide computing tools which model the various phases of severe accidents and will be validated on experiments performed with real and simulating materials. The main lines of the program are: - the development of the TOLBIAC 3D code for the thermal hydraulics of core melt pools, which will be validated against the Bali experiment presently under construction - the Sultan experiment, to study the capability of cooling by external flooding of the reactor vessel - the development of the MC-3D code for core melt

  4. 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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Shibata, Toshikazu.

    1982-01-01

    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 U 3 O 8 -Al fuel with high uranium densities, and thicker fuel meat - resulted in higher maximum fuel and surface temperatures with the LEU oxide fuel. (author)

  7. Thermal-hydraulic transient characteristics of ship-propulsion reactor investigated through safety analysis

    International Nuclear Information System (INIS)

    Fujiki, Kazuo; Asaka, Hideaki; Ishida, Toshihisa

    1986-01-01

    Thermal-hydraulic behaviors in the reactor of Nuclear Ship ''Mutsu'' were investigated through safety evaluation of operational transients by using RETRAN and COBRA-IV codes. The results were compared to the transient behaviors of typical commercial PWR and the characteristics of transient thermal-hydraulic behaviors in ship-loaded reactor were figured out. ''Mutsu'' reactor has larger thermal margin than commercial PWR because it is designed to be used as ship-propulsion power source in the load-following operation mode. This margin makes transient behavior in general milder than in commercial PWR but high opening pressure set point of main-steam safety valves leads poor heat-sink condition after reactor trip. The effects of other small-sized components are also investigated. The findings in the paper will be helpful in the design of future advanced reactor for nuclear ship. (author)

  8. Single-channel model for steady thermal-hydraulic analysis in nuclear reactor

    International Nuclear Information System (INIS)

    Zhang Xiaoying; Huang Yuanyuan

    2010-01-01

    This article established a single-channel model for steady analysis in the reactor and an example of thermal-hydraulic analysis was made by using this model, including the Maximum heat flux density of fuel element, enthalpy, Coolant flow, various kinds of pressure drop, enthalpy increase in average tube and thermal tube. I also got the Coolant temperature distribution and the fuel element temperature distribution and analysis of the final result. The results show that some relevant parameters which we got in this paper are well coincide with the actual operating parameters. It is also show that the single-channel model can be used to the steady thermal-hydraulic analysis. (authors)

  9. Advanced methodology to simulate boiling water reactor transient using coupled thermal-hydraulic/neutron-kinetic codes

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, Christoph Oliver

    2016-06-13

    Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools. The generation of cross-section (XS) libraries, depending on the individual thermal-hydraulic state parameters, is of paramount importance for coupled simulations. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running commercial and user-friendly lattice codes such as CASMO and HELIOS. In this dissertation a computational route, based on the lattice code SCALE6/TRITON, the cross-section interface GenPMAXS, the best-estimate thermal-hydraulic system code TRACE and the core simulator PARCS, for best-estimate simulations of Boiling Water (BWR) transients has been developed and validated. The computational route has been supplemented by a subsequent uncertainty and sensitivity study based on Monte Carlo sampling and propagation of the uncertainties of input parameters to the output (SUSA code). The analysis of a single BWR fuel assembly depletion problem with PARCS using SCALE/TRITON cross-sections has been shown a good agreement with the results obtained with CASMO cross-section sets. However, to compensate the deficiencies of the interface program GenPMAXS, PYTHON scripts had to be developed to incorporate missing data, as the yields of Iodine, Xenon and Promethium, into the cross-section-data sets (PMAXS-format) generated by GenPMAXS from the SCALE/TRITON output. The results of the depletion analysis of a full BWR core with PARCS have indicated the importance of considering history effects, adequate modeling of the reflector region and the control rods, as the PARCS simulations for depleted fuel and all control rods inserted (ARI) differs significantly at the fuel assembly top and bottom. Systematic investigations with the coupled codes TRACE/PARCS have been performed to analyse the core behaviour at different thermal conditions using nuclear data (XS

  10. Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

    OpenAIRE

    Itamar Iliuk; José Manoel Balthazar; Ângelo Marcelo Tusset; José Roberto Castilho Piqueira

    2016-01-01

    Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was prop...

  11. Thermal Performance Benchmarking: Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Feng, Xuhui [National Renewable Energy Laboratory (NREL), Golden, CO (United States). Transportation and Hydrogen Systems Center

    2017-10-19

    In FY16, the thermal performance of the 2014 Honda Accord Hybrid power electronics thermal management systems were benchmarked. Both experiments and numerical simulation were utilized to thoroughly study the thermal resistances and temperature distribution in the power module. Experimental results obtained from the water-ethylene glycol tests provided the junction-to-liquid thermal resistance. The finite element analysis (FEA) and computational fluid dynamics (CFD) models were found to yield a good match with experimental results. Both experimental and modeling results demonstrate that the passive stack is the dominant thermal resistance for both the motor and power electronics systems. The 2014 Accord power electronics systems yield steady-state thermal resistance values around 42- 50 mm to the 2nd power K/W, depending on the flow rates. At a typical flow rate of 10 liters per minute, the thermal resistance of the Accord system was found to be about 44 percent lower than that of the 2012 Nissan LEAF system that was benchmarked in FY15. The main reason for the difference is that the Accord power module used a metalized-ceramic substrate and eliminated the thermal interface material layers. FEA models were developed to study the transient performance of 2012 Nissan LEAF, 2014 Accord, and two other systems that feature conventional power module designs. The simulation results indicate that the 2012 LEAF power module has lowest thermal impedance at a time scale less than one second. This is probably due to moving low thermally conductive materials further away from the heat source and enhancing the heat spreading effect from the copper-molybdenum plate close to the insulated gate bipolar transistors. When approaching steady state, the Honda system shows lower thermal impedance. Measurement results of the thermal resistance of the 2015 BMW i3 power electronic system indicate that the i3 insulated gate bipolar transistor module has significantly lower junction

  12. Performance of nano-hydraulic turbine utilizing waterfalls

    Energy Technology Data Exchange (ETDEWEB)

    Ikeda, Toshihiko; Iio, Shouichiro; Tatsuno, Kenji [Faculty of Engineering, Shinshu University, 4-17-1 Wakasato, Nagano, 380-8553 (Japan)

    2010-01-15

    The aim of this investigation was to develop an environmentally friendly nano-hydraulic turbine utilizing waterfalls. A model of an impulse type hydraulic turbine constructed and tested with an indoor type waterfall to arrive at an optimum installation condition. Effects of an installation parameter, namely distance between the rotor and the waterfall on the power performance were studied. The flow field around the rotor was examined visually to clarify influences of installation conditions on the flow field. The flow visualization showed differences of flow pattern around the rotor by the change of flow rate and rotational speed of the rotor. From this study it was found that the power performances of the rotor were changed with the distance between the rotor and the waterfalls. The maximum power coefficient of this turbine is approximately 60%. Also, to respond to changes in the waterfall flow rate, we placed a flat plate on the upper side of the rotor to control the water flow direction. As a result, we found that the coefficient of this turbine is increased with the flow rate and power could be obtained even when the flow rate changed by 3.5 times if the plate was placed on the upper side of the rotor. Although the power coefficient decreased when the plate was installed, the power coefficient still is from 53 to 58%. (author)

  13. Techniques for the thermal/hydraulic analysis of LMFBR check valves