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Sample records for thermal hydraulic test

  1. 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.

  2. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    W.E. Lowry

    2001-12-13

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

  4. 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).

  5. 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

  6. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tadashi; Itoh, Hideo; Kiuchi, Toshio; Watanabe, Hironori; Kimura, Mamoru; Anoda, Yoshinari [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    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)

  8. Development of coupled neutronics/thermal-hydraulics test case for HPLWR

    Science.gov (United States)

    Pham, P.; Gamtsemlidze, I. D.; Bahdanovich, R. B.; Nikonov, S. P.; Smirnov, A. D.

    2017-01-01

    The High-Performance Light Water Reactor (HPLWR) is the European concept of a supercritical water reactor (SCWR) which is one of the most promising and innovative designs of the Generation IV nuclear reactor concepts. The thermal-hydraulics behavior of supercritical water is significantly different from water at sub-critical pressure because of the difference in the specific heat value. Coupled analysis of HPLWR assembly neutronics and thermal-hydraulics has become important because of the strong influence of the water density on the neutron spectrum and power distribution. Programs MCU (Monte-Carlo Universal) and ATHLET (Analysis of Thermal-hydraulics of Leaks and Transients) were used for better estimation of power and temperature distribution in HPLWR assembly.

  9. Thermal-Hydraulics and Electrochemistry of a Boiling Solution in a Porous Sludge Pile A Test Methodology

    Energy Technology Data Exchange (ETDEWEB)

    R.F. Voelker

    2001-05-03

    When boiling occurs in a pile of porous corrosion products (sludge), chemical species can concentrate. These species can react with the corrosion products and transform the sludge into a rock hard mass and/or create a corrosive environment. In-situ measurements are required to improve the understanding of this process, and the thermal-hydraulic and electrochemical environment in the pile. A test method is described that utilizes a water heated instrumented tube array in an autoclave to perform the in-situ measurements. As a proof of method feasibility, tests were performed in an alkaline phosphate solution. The test data is discussed. Temperature changes and electrochemical potential shifts were used to indicate when chemicals concentrate and if/when the pile hardens. Post-test examinations confirmed hardening occurred. Experiments were performed to reverse the hardening process. A one-dimensional model, utilizing capillary forces, was developed to understand the thermal-hydraulic measurements.

  10. Thermal-hydraulic analysis for changing feedwater check valve leakage rate testing methodology

    Energy Technology Data Exchange (ETDEWEB)

    Fuller, R.; Harrell, J.

    1996-12-01

    The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. This degraded performance was exhibited by frequent seal failures and subsequent valve repairs. The original requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leak path exists through the feedwater lines during the reactor blowdown phase and that sufficient subcooled water remains in various portions of the feedwater piping to form liquid water loop seals that effectively isolate this leak path. These results provided the bases for changing the leak testing requirements of the FWCVs from air to water. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves.

  11. Improvement of auditing technology of safety analysis through thermal-hydraulic separate effect tests

    Energy Technology Data Exchange (ETDEWEB)

    No, Hee Cheon; Moon, Young Min; Lee, Dong Won; Lee, Sang Ik; Kim, Eung Soo; Yeom, Keum Soo [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2002-03-15

    The objective of the present research is to perform the separate effect tests and to assess the RELAP5/MOD3.2 code for the analysis of thermal-hydraulic behavior in the reactor coolant system and the improvement of the auditing technology of safety analysis. Three Separate Effect Tests (SETs) are the reflux condensation in the U-tube, the direct contact condensation in the hot-leg and the mixture level buildup in the pressurizer. The experimental data and the empirical correlations are obtained through SETs. On the ases of the three SET works, models in RELAP5 are modified and improved, which are compared with the data. The Korea Standard Nuclear Power Plant (KSNP) are assessed using the modified RELAP5. In the reflux condensation test, the data of heat transfer coefficients and flooding are obtained and the condensation models are modified using the non-iterative model, as results, modified code better predicts the data. In the direct contact condensation test, the data of heat transfer coefficients are obtained for the cocurrent and countercurrent flow between the mixture gas and the water in condition of horizontal stratified flow. Several condensation and friction models are modified, which well predict the present data. In the mixture level test, the data for the mixture level and the onset of water draining into the surge line are obtained. The standard RELAP5 over-predicts the mixture level and the void fraction in the pressurizer. Simple modification of model related to the pool void fraction is suggested. The KSNP is assessed using the standard and the modified RELAP5 resulting from the experimental and code works for the SETs. In case of the pressurizer manway opening with available secondary side of the steam generators, the modified code predicts that the collapsed level in the pressurizer is little accumulated. The presence and location of the opening and the secondary condition of the steam generators have an effect on the coolant inventory. The

  12. Improvement of auditing technology of safety analysis through thermal-hydraulic separate effect tests

    Energy Technology Data Exchange (ETDEWEB)

    No, Hee Cheon; Park, Hyun Sik; Kim, Hyoung Tae; Moon, Young Min; Choi, Sung Won; Hwang, Do Hyun [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2000-03-15

    The direct-contact condensation hear transfer coefficients are experimentally obtained in the following conditions : pure steam/steam in the presence of noncondensible gas, horizontal/slightly inclined pipe, cocurrent/countercurrent stratified flow with water. The empirical correlation for liquid Nusselt number is developed in conditions of the slightly inclined pipe and the cocurrent stratified flow. The several models - the wall friction coefficient, the interfacial friction coefficient, the correlation of direct-contact condensation with noncondensible gases, and the correlation of wall film condensation - in the RELAP5/MOD3.2 code are modified, As results, RELAP5/MOD3.2 is improved. The present experimental data is used for evaluating the improved code. The standard RELAP5/MOD3.2 code is modified using the non-iterative modeling, which is a mechanistic model and does not require any interfacial information such as the interfacial temperature, The modified RELAP5/MOD3.2 code os used to simulate the horizontally stratified in-tube condensation experiment which represents the direct-contact condensation phenomena in a hot leg of a nuclear reactor. The modeling capabilities of the modified code as well as the standard code are assessed using several hot-leg condensation experiments. The modified code gives better prediction over local experimental data of liquid void fraction and interfacial heat transfer coefficient than the standard code. For the separate effect test of the thermal-hydraulic phenomena in the pressurizer, the scaling analysis is performed to obtain a similarity of the phenomena between the Korea Standard Nuclear Power Plant(KSNPP) and the present experimental facility. The diameters and lengths of the hot-leg, the surge line and the pressurizer are scaled down with the similitude of CCFL and velocity. The ratio of gas flow rate is 1/25. The experimental facility is composed of the air-water supply tank, the horizontal pipe, the surge line and the

  13. ATHENA (Advanced Thermal-Hydraulic Energy Network Analyzer) transient analysis of a fusion engineering test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wareing, T.A.

    1988-02-01

    Two potential undercooling transients are of concern in the design of TIBER-II (Tokamak Ignition/Burn Experimental Reactor), namely loss of coolant and loss of flow accidents. The major area of concern for TIBER-II is the inboard shield, where, due to tungsten material, the decay heat is extremely high. The purpose of this study was to analyze these transients using the thermal-hydraulic code ATHENA (Advanced Thermal-Hydraulic Energy Network Analyzer). The most comprehensive portion of this project involved creating a simple, yet complete, ATHENA model representative of TIBER-II. The completed model represents the case when the plasma is off and contains the inboard shield, the outboard shield, the divertor shields, and the primary loop. The primary loop contains the piping, pump, pressurizer, and heat exchanger. The heat exchanger is at the same elevation as the reactor, the least favorable to establishing natural circulation. The only transient analyzed so far, however, is a loss of flow accident. Results from the loss of flow analysis show that there is sufficient natural circulation in the inboard, outboard, and lower divertor shield to remove the decay heat, assuming that the secondary side flow is at full capacity. Although the upper divertor shield does not have sufficient natural circulation, cooling is provided due to vaporization and re-flood oscillations. However, one must recognize that there may be some local hot sport where the flow geometry inhibits cooling in a LOFA; the ATHENA model would not detect any localized problem. 9 refs., 18 figs., 4 tabs.

  14. Thermal hydraulic analysis of China fusion engineering test reactor during thermal quenching by comparative approach of Relap5 and THEATRe codes

    Science.gov (United States)

    Khan, Salah Ud-Din; Song, Yuntao; Khan, Shahab Ud-Din

    2016-10-01

    Thermal quenching in Tokamak reactor is the most obvious phenomenon happens during plasma disruption conditions. The current research is focused on the thermal behavior of different parameters of China fusion engineering test reactor (CFETR) including reactor power, pressure and mass flow rate conditions. The analysis was performed by two thermal hydraulic codes, i.e. THEATRe and Relap5. During the first phase of research, thermal quenching behavior and trends that can be possible during the reactor operation was performed. In the next phase, nodalization diagram of THEATRe and Relap5 codes were developed. The listed parameters were calculated and analyzed for the safety aspects of the reactor. The main objective of the research was to analyze the blanket system of CFETR (Tokamak) for safety concerns during disruption condition. The research will be extended to other components for safe operation of reactor as well.

  15. Integral and Separate Effects Tests for Thermal Hydraulics Code Validation for Liquid-Salt Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per

    2012-10-30

    The objective of the 3-year project was to collect integral effects test (IET) data to validate the RELAP5-3D code and other thermal hydraulics codes for use in predicting the transient thermal hydraulics response of liquid salt cooled reactor systems, including integral transient response for forced and natural circulation operation. The reference system for the project is a modular, 900-MWth Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a specific type of Fluoride salt-cooled High temperature Reactor (FHR). Two experimental facilities were developed for thermal-hydraulic integral effects tests (IETs) and separate effects tests (SETs). The facilities use simulant fluids for the liquid fluoride salts, with very little distortion to the heat transfer and fluid dynamics behavior. The CIET Test Bay facility was designed, built, and operated. IET data for steady state and transient natural circulation was collected. SET data for convective heat transfer in pebble beds and straight channel geometries was collected. The facility continues to be operational and will be used for future experiments, and for component development. The CIET 2 facility is larger in scope, and its construction and operation has a longer timeline than the duration of this grant. The design for the CIET 2 facility has drawn heavily on the experience and data collected on the CIET Test Bay, and it was completed in parallel with operation of the CIET Test Bay. CIET 2 will demonstrate start-up and shut-down transients and control logic, in addition to LOFC and LOHS transients, and buoyant shut down rod operation during transients. Design of the CIET 2 Facility is complete, and engineering drawings have been submitted to an external vendor for outsourced quality controlled construction. CIET 2 construction and operation continue under another NEUP grant. IET data from both CIET facilities is to be used for validation of system codes used for FHR modeling, such as RELAP5-3D. A set of

  16. 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.

  17. 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.

  18. Assessment of the Thermal Hydraulic Models in THALES

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Byeong Il; Kim, Hong Ju; Jang, Beomjun; Woo, Hae-Seuk [KEPCO Nuclear Fuel, Daejeon (Korea, Republic of)

    2016-10-15

    THALES (Thermal Hydraulic AnaLyzer for Enhanced Simulation of core) developed by KEPCO Nuclear Fuel is a subchannel analysis code on the basis of the single-stage core analysis model. THALES calculates the local fluid conditions and DNBR in the PWR (Pressurized Water Reactor) core. Currently, THALES is limited to the licensed type of the nuclear power plant because the thermal hydraulic models and CHF (Critical Heat Flux) correlations for OPR1000 and APR1400 are only licensed. KEPCO NF intends to apply THALES to WH typed nuclear power plants in Korea. To expand the applicable types of the nuclear power plants, the existing thermal hydraulic models were modified and new thermal hydraulic models were added to THALES. In this study, the thermal hydraulic models tested and added in THALES are reviewed and a preliminary calculation is performed. KEPCO NF intends to apply THALES to various typed nuclear power plants in Korea. So, the existing thermal hydraulic models implemented in THALES are modified and the void model which are generally used in the subchannel analysis code is added. Through the preliminary calculation, it is confirmed that the thermal hydraulic models are properly modified and implemented in THALES, which shows the possibility to apply THALES in various typed nuclear power plants in Korea.

  19. Thermal hydraulic tests for developing two-phase thermo-siphon loop of CARR-CNS

    Science.gov (United States)

    Du, Shejiao; Bi, Qincheng; Chen, Tingkuan; Feng, Quanke

    2005-04-01

    The China Institute of Atomic Energy (CIAE) is now constructing the China Advanced Research Reactor (CARR: 60 MW), and designing the cold neutron source (CNS) with a two-phase hydrogen thermo-siphon loop consisting of a condenser, a single-moderator transfer tube and a cylindrical annulus moderator cell. The mockup tests reported here were carried out on a full-scale loop using Freon-113 as the working fluid in order to validate the self-regulating characteristics of the loop, with a void fraction less than 20% in the liquid of the moderator cell and requirements for establishing the condition under which the inner shell of the cell has only vapor and the outer shell liquid. During these mockup tests, the density ratio of liquid to vapor and the vapor volumetric evaporation rate due to heat load were kept the same as those in normal operation of the CARR-CNS. The results show that the loop has the required self-regulating characteristics and the inner shell of the moderator cell contains only vapor, the outer shell liquid. The local void fractions in the liquid increase with an increase in loop pressure under the condition of a constant vapor volumetric evaporation rate.

  20. 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.

  1. Improvement of auditing technology of safety analysis through thermal-hydraulic separate effect tests

    Energy Technology Data Exchange (ETDEWEB)

    No, Hee Cheon; Park, Hyun Sik; Kim, Hyougn Tae; Moon, Young Min; Choi, Sung Won; Heo, Sun [Korea Advanced Institute Science and Technology, Taejon (Korea, Republic of)

    1999-04-15

    The loss-of-RHR accident during midloop operation has been important as results of the probabilistic safety analysis. The condensation models In RELAP5/MOD3 are not proper to analyze the midloop operation. To audit and improve the model in RELAP5/MOD3.2, several items of separate effect tests have been performed. The 29 sets of reflux condensation data is obtained and the correlation is developed with these heat transfer coefficient's data. In the experiment of the direct contact condensation in hot leg, the apparatus setting is finished and a few experimental data is obtained. Non-iterative model is used to predict the model in RELAP5/MOD3.2 with the results of reflux condensation and evaluates better than the present model. The results of the direct contact condensation in a hot leg represent to be similar with the present model. The study of the CCF and liquid entrainment in a surge line and pressurizer is selected as the third separate experiment and is on performance.

  2. Gas Test Loop Booster Fuel Hydraulic Testing

    Energy Technology Data Exchange (ETDEWEB)

    Gas Test Loop Hydraulic Testing Staff

    2006-09-01

    The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3.

  3. Feasibility Test of Commercial CFD Code for Thermal-Hydraulic Analysis of Wire-wrapped Fuel Pin Bundle in SFR

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chiwoong; Ha, Kwiseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Haeyong, Jeong [Sejong Univ., Daejeon (Korea, Republic of)

    2014-05-15

    One of the major difficulties in the numerical approach is a grid generation of the interface between a wire and rod, because the point connectivity and higher scale deviation of a wire and rod can easily increase the number of meshes. The general way to handle this problem is a simplification of the wire-rod interface. In this study, the wire configuration effect is also studied with a no-wire case and simplified wire geometries. The materials of coolant and cladding are sodium and HT9. Fuel part is not modeled. The detailed design parameters are described in Table 1. In addition, Fig. 2 shows a detailed schematic of the fuel rod and wire. The wire contact distance, Sw, is the length penetrating the clad. Basically, the wire will be firmly contacted with the clad surface. As shown in Table 1, the wire diameter is the same as the gap between two fuel rods. However, it is unavoidable in the modeling of the wire to ignore a singular contact point between the wire and cladding. Therefore, an intersected wire and rod are arbitrarily generated. The boundary conditions for the conjugated heat transfer analysis are described in Table 2. The inlet region is defined with a uniform velocity and constant temperature of 650 K. The outlet is defined with a constant ambient pressure, i. e., zero static pressure. Since only the cladding part was modeled, a uniform heat flux condition is applied on the inner cladding wall surface. All solid surfaces are considered as no slip adiabatic boundaries. The numerical analyses of wire-wrapped 7 pins are conducted using ANSYS-CFX to check the feasibility of the CFD tool in thermal-hydraulic phenomena in a wire-wrapped fuel pin bundle. Typical turbulent models are applied to check the dependency of model selection on the pressure drop and heat transfer on the wire-wrapped fuel rod. In short, the omega-based model shows the highest pressure drop and heat transfer. Comparing the existing correlations, the pressure drop results represent

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-06-30

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

  5. 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 ...

  6. Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    C. B. Davis

    2006-07-01

    The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code’s calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.

  7. 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.

  8. Thermal-Hydraulic Integral Effect Test with ATLAS for an Intermediate Break Loss of Coolant Accident at a Pressurizer Surge Line

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Seok Cho; Park, Hyun Sik; Choi, Nam Hyun; Park, Yu Sun; Kim, Jong Rok; Bae, Byoung Uhn; Kim, Yeon Sik; Kim, Kyung Doo; Choi, Ki Yong; Song, Chul Hwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The main objectives of this test were not only to provide physical insight into the system response of the APR1400 during the pressurizer surge line break accident but also to produce an integral effect test data to validate the SPACE code. In order to simulate a double-ended guillotine break of a pressurizer surge line in the APR1400, the IB-SUR-01R test was performed with ATLAS. The major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Despite the core was uncovered, no excursion in the cladding temperature was observed. The pressurizer surge line break can be classified as a hot leg break from a break location point of view. Compared with a cold leg break, coolability in the core may be better in case of a hot leg break due to the enhanced flow in the core region. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code. Furthermore, this data can be utilized to identify any code deficiency for an IBLOCA simulation, especially for DVI-adapted plants. Redefinition of break size for design basis accident (DBA) based on risk information is being extensively investigated due to the potential for safety benefits and unnecessary burden reduction from current LBLOCA (large break loss of coolant accident)-based ECC (Emergency Core Cooling) Acceptance Criteria. As a transition break size (TBS), the rupture of medium-size pipe is considered to be more important than ever in risk-informed regulation (RIR)-relevant safety analysis. As plants age, are up-rated, and continue to seek improved operating efficiencies, the small break and intermediate break LOCA (IBLOCA) can become a concern. In particular, IBLOCA with DVI (Direct Vessel Injection) features will be addressed to support redefinition of a design-basis LOCA. With an aim of expanding code validation to address small

  9. Numerical simulation of the direct contact condensation phenomena for PTS-related in single and combined-effect thermal hydraulic test facilities using TransAT CMFD code

    Energy Technology Data Exchange (ETDEWEB)

    Kadi, Rabah, E-mail: kadi.rkhaled@hotmail.com [Laboratory for Thermal-Hydraulics, Nuclear Research Center of Birine (Algeria); Aissani, Slimane [Hydrocarbons and Chemistry Faculty, University of Boumerdes (Algeria); Bouam, Abdellah [Laboratory for Thermal-Hydraulics, Nuclear Research Center of Birine (Algeria)

    2015-11-15

    Highlights: • TransAT CMFD code application to DCC phenomenon. • LEIS methodology to predict the condensing steam flow rate. • Validation of interfacial phase-change heat transfer and turbulence models. • Correction of damping function at the free surface region. • Numerical validation of previous models using LIM and KAERI & KAIST test facilities. - Abstract: The use of CFD for the industrial studies related to PTS, including DCC is already possible; improvements of the two-phase modeling capabilities have to be undertaken to qualify the codes for the simulation of such flows. The DCC in horizontally stratified flow regime constitutes very considerable challenge exercises for a computational fluid dynamics (CFD) simulation of the thermal hydraulics PTS phenomenon because the interplay between turbulence and interfacial heat and mass transfer problem. The main purpose of our study is to investigate numerically the DCC in horizontally stratified steam water flow in a 2D and 3D channel using TransAT CMFD code. The new methodology known as Large-Eddy & Interface (LEIS) have been implemented for treatment of turbulence combined with interface tracking ITM (level set approach). Among of the so-called ‘coarse-grained’ ITM's models, the modified original surface divergence has been chosen as well as the treatment of the turbulence by URANS and VLES. This contribution addressed on the validation of interfacial phase-change heat transfer and turbulence models with special correction of the damping function at the free surface for single and combined-effect thermal hydraulic studies for LIM and KAERI & KAIST test facilities. The LIES methodology was found to apply successfully to predict the condensing steam flow rate in the all cases of the LIM test case involving a Smooth to Wavy turbulent, concurrent stratified steam-water flow in a 2D channel. The CMFD TransAT code predicting capability is analyzed, comparing the liquid temperature and to much the

  10. 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...

  11. 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...

  12. 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.

  13. Benchmark Simulations of the Thermal-Hydraulic Responses during EBR-II Inherent Safety Tests using SAM

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Rui; Sumner, Tyler S.

    2016-04-17

    An advanced system analysis tool SAM is being developed for fast-running, improved-fidelity, and whole-plant transient analyses at Argonne National Laboratory under DOE-NE’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. As an important part of code development, companion validation activities are being conducted to ensure the performance and validity of the SAM code. This paper presents the benchmark simulations of two EBR-II tests, SHRT-45R and BOP-302R, whose data are available through the support of DOE-NE’s Advanced Reactor Technology (ART) program. The code predictions of major primary coolant system parameter are compared with the test results. Additionally, the SAS4A/SASSYS-1 code simulation results are also included for a code-to-code comparison.

  14. Development of a kinetics analysis code for fuel solution combined with thermal-hydraulics analysis code PHOENICS and analysis of natural-cooling characteristic test of TRACY. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Shouichi; Yamane, Yuichi; Miyoshi, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Since exact information is not always acquired in the criticality accident of fuel solution, parametric survey calculations are required for grasping behaviors of the thermal-hydraulics. On the other hand, the practical methods of the calculation with can reduce the computation time with allowable accuracy will be also required, since the conventional method takes a long calculation time. In order to fulfill the requirement, a two-dimensional (R-Z) nuclear-kinetics analysis code considering thermal-hydraulic based on the multi-region kinetic equations with one-group neutron energy was created by incorporating with the thermal-hydraulics analysis code PHOENICS for all-purpose use the computation time of the code was shortened by separating time mesh intervals of the nuclear- and heat-calculations from that of the hydraulics calculation, and by regulating automatically the time mesh intervals in proportion to power change rate. A series of analysis were performed for the natural-cooling characteristic test using TRACY in which the power changed slowly for 5 hours after the transient power resulting from the reactivity insertion of a 0.5 dollar. It was found that the code system was able to calculate within the limit of practical time, and acquired the prospect of reproducing the experimental values considerably for the power and temperature change. (author)

  15. 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.

  16. Computer code analysis of steam generator in thermal-hydraulic test facility simulating nuclear power plant; Ydinvoimalaitosta kuvaavan koelaitteiston hoeyrystimien analysointi tietokoneohjelmilla

    Energy Technology Data Exchange (ETDEWEB)

    Virtanen, E.

    1995-12-31

    In the study three loss-of-feedwater type experiments which were preformed with the PACTEL facility has been calculated with two computer codes. The purpose of the experiments was to gain information about the behaviour of horizontal steam generator in a situation where the water level on the secondary side of the steam generator is decreasing. At the same time data that can be used in the assessment of thermal-hydraulic computer codes was assembled. The purpose of the work was to study the capabilities of two computer codes, APROS version 2.11 and RELAP5/MOD3.1, to calculate the phenomena in horizontal steam generator. In order to make the comparison of the calculation results easier the same kind of model of the steam generator was made for both codes. Only the steam generator was modelled, the rest of the facility was given for the codes as a boundary condition. (23 refs.).

  17. Data Collecting and Processing System and Hydraulic Control System of Hydraulic Support Model Test

    Directory of Open Access Journals (Sweden)

    Hong-Yu LIU

    2014-10-01

    Full Text Available Hydraulic support is an important equipment of mechanization caving coal in modernization coal mine. Hydraulic support must pass national strength test before it quantity production and use. Hydraulic support model test based on similarity theory is a new effective hydraulic support design and test method. The test information such as displacement, stress, strain and so on can be generalized to hydraulic support prototype, which can prompt hydraulic support design. In order to satisfy the need of hydraulic support model test, the data collecting and processing system of hydraulic support model test was established, relative software was programmed, the tress computation software of practical measurement data of hydraulic support model test was programmed, which provide practical and convenient research method for hydraulic support model test. By the data collecting and processing system software of hydraulic support model test and related software, user can realize the function such as data collecting, real time display, saving, analysis and processing to strain signals. The construction of load equipment and hydraulic control system of hydraulic support model test provides a practical and convenient research way for hydraulic support model test.

  18. 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.

  19. 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.

  20. 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...

  1. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Corradin, Michael [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Anderson, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Muci, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Dominguez, A. [Texas A & M Univ., College Station, TX (United States); Tokuhiro, Akira [Univ. of Idaho, Moscow, ID (United States); Hamman, K. [Univ. of Idaho, Moscow, ID (United States)

    2014-10-15

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  2. PMK-2, the First Integral Thermal-Hydraulics Tests for the Safety Evaluation of VVER-440/213 Nuclear Power Plants

    Directory of Open Access Journals (Sweden)

    Gy. Ézsöl

    2012-01-01

    Full Text Available The PMK-2 facility is a full-pressure thermal-hydraulic model of the primary and partly the secondary circuit of the VVER-type units of Paks NPP. The facility was the first integral-type facility for VVERs. The PMK-2 was followed later by the PACTEL (for VVER-440, the ISB, and PSB for VVER-1000. Since the startup of the facility in 1985, 55 experiments have been performed primarily in international frameworks with the participation of experts from 29 European and overseas countries forming a scientific school to better understand VVER system behaviour and reach a high level of modelling of accident sequences. The ATHLET, CATHARE, and RELAP5 codes have been validated including both qualitative and quantitative assessments. The former was almost exclusively applied to the early phase of validation by integral experiments, while the quantitative assessments have been performed by the Fast Fourier Transform Based Method. Paper gives comprehensive information on the design features of PMK-2 facility with a special respect to the representativeness of phenomena, the experiments performed, and the results of the validation of ATHLET, CATHARE, and RELAP5 codes. Safety significance of the PMK-2 projects is also discussed.

  3. 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.

  4. Application of computational fluid dynamics methods to improve thermal hydraulic code analysis

    Science.gov (United States)

    Sentell, Dennis Shannon, Jr.

    A computational fluid dynamics code is used to model the primary natural circulation loop of a proposed small modular reactor for comparison to experimental data and best-estimate thermal-hydraulic code results. Recent advances in computational fluid dynamics code modeling capabilities make them attractive alternatives to the current conservative approach of coupled best-estimate thermal hydraulic codes and uncertainty evaluations. The results from a computational fluid dynamics analysis are benchmarked against the experimental test results of a 1:3 length, 1:254 volume, full pressure and full temperature scale small modular reactor during steady-state power operations and during a depressurization transient. A comparative evaluation of the experimental data, the thermal hydraulic code results and the computational fluid dynamics code results provides an opportunity to validate the best-estimate thermal hydraulic code's treatment of a natural circulation loop and provide insights into expanded use of the computational fluid dynamics code in future designs and operations. Additionally, a sensitivity analysis is conducted to determine those physical phenomena most impactful on operations of the proposed reactor's natural circulation loop. The combination of the comparative evaluation and sensitivity analysis provides the resources for increased confidence in model developments for natural circulation loops and provides for reliability improvements of the thermal hydraulic code.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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)

  6. Development of fuel performance and thermal hydraulic technology

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-01

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

  7. HELOKA-HP thermal-hydraulic model validation and calibration

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Xue Zhou; Ghidersa, Bradut-Eugen; Badea, Aurelian Florin

    2016-11-01

    Highlights: • The electrical heater in HELOKA-HP has been modeled with RELAP5-3D using experimental data as input. • The model has been validated using novel techniques for assimilating experimental data and the representative model parameters with BEST-EST. • The methodology is successfully used for reducing the model uncertainties and provides a quantitative measure of the consistency between the experimental data and the model. - Abstract: The Helium Loop Karlsruhe High Pressure (HELOKA-HP) is an experimental facility for the testing of various helium-cooled components at high temperature (500 °C) and high pressure (8 MPa) for nuclear fusion applications. For modeling the loop thermal dynamics, a thermal-hydraulic model has been created using the system code RELAP5-3D. Recently, new experimental data covering the behavior of the loop components under relevant operational conditions have been made available giving the possibility of validating and calibrating the existing models in order to reduce the uncertainties of the simulated responses. This paper presents an example where such process has been applied for the HELOKA electrical heater model. Using novel techniques for assimilating experimental data, implemented in the computational module BEST-EST, the representative parameters of the model have been calibrated.

  8. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Heard, F.J.; Cramer, E.R.; Beaver, T.R. [Westinghouse Hanford Co., Richland, WA (United States); Thurgood, M.J. [Marvin (John), Inc. (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Thomas Schulenberg; Xu Cheng; Robert Stieglitz [Institute for Nuclear and Energy Technologies, Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2005-07-01

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

  12. 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.

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

    Directory of Open Access Journals (Sweden)

    Khater Hany

    2006-01-01

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

  14. 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.

  15. On the possible role of thermal radiation in containment thermal–hydraulics experiments by the example of CFD analysis of TOSQAN T114 air–He test

    Energy Technology Data Exchange (ETDEWEB)

    Filippov, A.S.; Grigoryev, S.Yu. [Nuclear Safety Institute of the Russian Academy of Sciences, Moscow (Russian Federation); Moscow Institute of Physics and Technology (Russian Federation); Tarasov, O.V. [Nuclear Safety Institute of the Russian Academy of Sciences, Moscow (Russian Federation)

    2016-12-15

    Highlights: • Neglecting by heat radiation in simulation of containment tests may cause discrepancies. • To show that, heat exchange in T114 air-helium test was analyzed in different ways. • Effect of thermal radiation on local temperature was numerically obtained in air with ∼1% steam content. • Model of gas-structure heat exchange in containment should include heat radiation. - Abstract: One of the experiments of ERCOSAM–SAMARA (E–S) projects (TOSQAN T114) is examined from the viewpoint of the radiative heat transfer (RHT) contribution to the overall heat exchange. E–S projects and T114 test were focused on investigation of light gas stratification in severe accident containment atmosphere and stratification break-up after the activation of mitigation systems. The first from two phases of T114 test is considered during which helium is quasistatically injected into the upper part of the TOSQAN vessel having isothermal walls and initially filled by air. The developing free convection removes most of the heat acquired, but not all. Thus stable local deviations in calculated temperatures were obtained in simulations that were interpreted as the deficiencies of the physical heat-transfer model. The modeling of RHT was included in full CFD simulation that resulted in a better agreement in local temperatures. The results of comparative calculations performed without/with RHT modeling are described in the paper. The RHT model implemented in the used CFD code (ANSYS FLUENT) was tested on known analytical solutions. The RHT contribution in T114 test was also estimated analytically to demonstrate independently that it may be noticeable in this experiment. The same estimations may be valid for stagnant zones of severe accident containment. All that shows the need in further detailing of the role of RHT in gas-structure heat exchange: as for interpretation of some containment tests performed in pressure vessel as for containment modeling.

  16. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-10-01

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

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

    Science.gov (United States)

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

    2007-01-01

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

  19. Ouellette Thermal Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Thermal Test Facility is a joint Army/Navy state-of-the-art facility (8,100 ft2) that was designed to:Evaluate and characterize the effect of flame and thermal...

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

    OpenAIRE

    MITRAN Tudor; CHIOREANU Nicolae; ABAITANCAI Horia; RUS Alexandru

    2016-01-01

    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.

  1. 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.

  2. 78 FR 8202 - Meeting of the Joint ACRS Subcommittees on Thermal Hydraulic Phenomena and Materials, Metallurgy...

    Science.gov (United States)

    2013-02-05

    ... Hydraulic Phenomena and Materials, Metallurgy and Reactor Fuels; Notice of Meeting The Joint ACRS Subcommittees on Thermal Hydraulic Phenomena and Materials, Metallurgy and Reactor Fuels will hold a meeting on...

  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-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.

  5. 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.

  6. Hydraulic properties from pumping tests data of aquifers in Azare ...

    African Journals Online (AJOL)

    Hydraulic properties from pumping tests data of aquifers in Azare area, North Eastern Nigeria. AC Tse, PA Amadi. Abstract. Pumping test data from twelve boreholes in Azare area were analysed to determine the hydraulic properties of the aquifers, and the availability of water to meet the conjugate demands of the increasing ...

  7. 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.

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

    Science.gov (United States)

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

    1988-01-01

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

  9. INFORMATION-MEASURING TEST SYSTEM OF DIESEL LOCOMOTIVE HYDRAULIC TRANSMISSIONS

    Directory of Open Access Journals (Sweden)

    I. V. Zhukovytskyy

    2015-08-01

    Full Text Available Purpose. The article describes the process of developing the information-measuring test system of diesel locomotives hydraulic transmission, which gives the possibility to obtain baseline data to conduct further studies for the determination of the technical condition of diesel locomotives hydraulic transmission. The improvement of factory technology of post-repair tests of hydraulic transmissions by automating the existing hydraulic transmission test stands according to the specifications of the diesel locomotive repair enterprises was analyzed. It is achieved based on a detailed review of existing foreign information-measuring test systems for hydraulic transmission of diesel locomotives, BelAZ earthmover, aircraft tug, slag car, truck, BelAZ wheel dozer, some brands of tractors, etc. The problem for creation the information-measuring test systems for diesel locomotive hydraulic transmission is being solved, starting in the first place from the possibility of automation of the existing test stand of diesel locomotives hydraulic transmission at Dnipropetrovsk Diesel Locomotive Repair Plant "Promteplovoz". Methodology. In the work the researchers proposed the method to create a microprocessor automated system of diesel locomotives hydraulic transmission stand testing in the locomotive plant conditions. It acts by justifying the selection of the necessary sensors, as well as the application of the necessary hardware and software for information-measuring systems. Findings. Based on the conducted analysis there was grounded the necessity of improvement the plant hydraulic transmission stand testing by creating a microprocessor testing system, supported by the experience of developing such systems abroad. Further research should be aimed to improve the accuracy and frequency of data collection by adopting the more modern and reliable sensors in tandem with the use of filtering software for electromagnetic and other interference. Originality. The

  10. 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.

  11. 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

  12. Coupled neutronics - thermal-hydraulics programs for SCWRS

    Energy Technology Data Exchange (ETDEWEB)

    Reiss, T. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Muegyetem rkp. 9., 1111 Budapest (Hungary)

    2010-07-01

    The Supercritical Water Cooled Reactor (SCWR) was chosen as one of the Generation IV reactors by GIF. At the moment, a number of concepts - thermal as well as fast ones - exist. The reference parameters for a thermal SCWR have been taken from the European High Performance Light Water Reactor (HPLWR). Since the pressure is higher than the critical pressure (22.1 MPa) there is no change in the phase of the water in the core. On the other hand, 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. This inclination is increased by the pseudo-critical transformation of the water used as coolant. Thus, for modelling a system of this type coupled neutronics - thermal-hydraulics programs are required. Such a program system has been developed with the following main features: great modularity which allows for easy modifications, thus several SCWR concepts can be studied; detailed assembly calculations (with MCNP) and full-core analysis (with SCALE) are supported; the differential equations of xenon poisoning are implemented to study xenon oscillations. The program system was used to examine the assembly of the HPLWR, to design the assembly and the core of the Simplified Supercritical Water Cooled Reactor (SSCWR) and to model xenon oscillations in SCWRs. (authors)

  13. Response of Compacted Bentonites to Thermal and Thermo-Hydraulic Loadings at High Temperatures

    Directory of Open Access Journals (Sweden)

    Snehasis Tripathy

    2017-07-01

    Full Text Available The final disposal of high-level nuclear waste in many countries is preferred to be in deep geological repositories. Compacted bentonites are proposed for use as the buffer surrounding the waste canisters which may be subjected to both thermal and hydraulic loadings. A significant increase in the temperature is anticipated within the buffer, particularly during the early phase of the repository lifetime. In this study, several non-isothermal and non-isothermal hydraulic tests were carried on compacted MX80 bentonite. Compacted bentonite specimens (water content = 15.2%, dry density = 1.65 Mg/m3 were subjected to a temperature of either 85 or 150 °C at one end, whereas the temperature at the opposite end was maintained at 25 °C. During the non-isothermal hydraulic tests, water was supplied from the opposite end of the heat source. The temperature and relative humidity were monitored along predetermined depths of the specimens. The profiles of water content, dry density, and degree of saturation were established after termination of the tests. The test results showed that thermal gradients caused redistribution of the water content, whereas thermo-hydraulic gradients caused both redistribution and an increase in the water content within compacted bentonites, both leading to development of axial stress of various magnitudes. The applied water injection pressures (5 and 600 kPa and temperature gradients appeared to have very minimal impact on the magnitude of axial stress developed. The thickness of thermal insulation layer surrounding the testing devices was found to influence the temperature and relative humidity profiles thereby impacting the redistribution of water content within compacted bentonites. Under the influence of both the applied thermal and thermo-hydraulic gradients, the dry density of the bentonite specimens increased near the heat source, whereas it decreased at the opposite end. The test results emphasized the influence of

  14. Effects of temperature and thermally-induced microstructure change on hydraulic conductivity of Boom Clay

    Directory of Open Access Journals (Sweden)

    W.Z. Chen

    2017-06-01

    Full Text Available Boom Clay is one of the potential host rocks for deep geological disposal of high-level radioactive nuclear waste in Belgium. In order to investigate the mechanism of hydraulic conductivity variation under complex thermo-mechanical coupling conditions and to better understand the thermo-hydro-mechanical (THM coupling behaviour of Boom Clay, a series of permeability tests using temperature-controlled triaxial cell has been carried out on the Boom Clay samples taken from Belgian underground research laboratory (URL HADES. Due to its sedimentary nature, Boom Clay presents across-anisotropy with respect to its sub-horizontal bedding plane. Direct measurements of the vertical (Kv and horizontal (Kh hydraulic conductivities show that the hydraulic conductivity at 80 °C is about 2.4 times larger than that at room temperature (23 °C, and the hydraulic conductivity variation with temperature is basically reversible during heating–cooling cycle. The anisotropic property of Boom Clay is studied by scanning electron microscope (SEM tests, which highlight the transversely isotropic characteristics of intact Boom Clay. It is shown that the sub-horizontal bedding feature accounts for the horizontal permeability higher than the vertical one. The measured increment in hydraulic conductivity with temperature is lower than the calculated one when merely considering the changes in water kinematic viscosity and density with temperature. The nuclear magnetic resonance (NMR tests have also been carried out to investigate the impact of microstructure variation on the THM properties of clay. The results show that heating under unconstrained boundary condition will produce larger size of pores and weaken the microstructure. The discrepancy between the hydraulic conductivity experimentally measured and predicted (considering water viscosity and density changes with temperature can be attributed to the microstructural weakening effect on the thermal volume change

  15. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Strizhov, V.; Kanukova, V.; Vinogradova, T.; Askenov, E. [Russian Academy of Sciences, Moscow (Russian Federation). Inst. of Nuclear Safety; Nikulshin, V. [Russian Research Center, Moscow (Russian Federation). Kurchatov Inst.

    1996-09-01

    This report deals with the subject of CORCON-Mod3 code validation (thermal-hydraulic modeling capability only) based on MCCI (molten core concrete interaction) experiments conducted under different programs in the past decade. Thermal-hydraulic calculations (i.e., concrete ablation, melt temperature, melt energy, concrete temperature, and condensible and non-condensible gas generation) were performed with the code, and compared with the data from 15 experiments, conducted at different scales using both simulant (metallic and oxidic) and prototypic melt materials, using different concrete types, and with and without an overlying water pool. Sensitivity studies were performed in a few cases involving, for example, heat transfer from melt to concrete, condensed phase chemistry, etc. Further, special analysis was performed using the ACE L8 experimental data to illustrate the differences between the experimental and the reactor conditions, and to demonstrate that with proper corrections made to the code, the calculated results were in better agreement with the experimental data. Generally, in the case of dry cavity and metallic melts, CORCON-Mod3 thermal-hydraulic calculations were in good agreement with the test data. For oxidic melts in a dry cavity, uncertainties in heat transfer models played an important role for two melt configurations--a stratified geometry with segregated metal and oxide layers, and a heterogeneous mixture. Some discrepancies in the gas release data were noted in a few cases.

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

    Directory of Open Access Journals (Sweden)

    Alessandro Petruzzi

    2008-01-01

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

  18. 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.

  19. 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

  20. Thermal Testing Measurements Report

    Energy Technology Data Exchange (ETDEWEB)

    R. Wagner

    2002-09-26

    The purpose of the Thermal Testing Measurements Report (Scientific Analysis Report) is to document, in one report, the comprehensive set of measurements taken within the Yucca Mountain Project Thermal Testing Program since its inception in 1996. Currently, the testing performed and measurements collected are either scattered in many level 3 and level 4 milestone reports or, in the case of the ongoing Drift Scale Test, mostly documented in eight informal progress reports. Documentation in existing reports is uneven in level of detail and quality. Furthermore, while all the data collected within the Yucca Mountain Site Characterization Project (YMP) Thermal Testing Program have been submitted periodically to the Technical Data Management System (TDMS), the data structure--several incremental submittals, and documentation formats--are such that the data are often not user-friendly except to those who acquired and processed the data. The documentation in this report is intended to make data collected within the YMP Thermal Testing Program readily usable to end users, such as those representing the Performance Assessment Project, Repository Design Project, and Engineered Systems Sub-Project. Since either detailed level 3 and level 4 reports exist or the measurements are straightforward, only brief discussions are provided for each data set. These brief discussions for different data sets are intended to impart a clear sense of applicability of data, so that they will be used properly within the context of measurement uncertainty. This approach also keeps this report to a manageable size, an important consideration because the report encompasses nearly all measurements for three long-term thermal tests. As appropriate, thermal testing data currently residing in the TDMS have been reorganized and reformatted from cumbersome, user-unfriendly Input-Data Tracking Numbers (DTNs) into a new set of Output-DTNs. These Output-DTNs provide a readily usable data structure

  1. Influence of pressure change during hydraulic tests on fracture aperture.

    Science.gov (United States)

    Ji, Sung-Hoon; Koh, Yong-Kwon; Kuhlman, Kristopher L; Lee, Moo Yul; Choi, Jong Won

    2013-03-01

    In a series of field experiments, we evaluate the influence of a small water pressure change on fracture aperture during a hydraulic test. An experimental borehole is instrumented at the Korea Atomic Energy Research Institute (KAERI) Underground Research Tunnel (KURT). The target fracture for testing was found from the analyses of borehole logging and hydraulic tests. A double packer system was developed and installed in the test borehole to directly observe the aperture change due to water pressure change. Using this packer system, both aperture and flow rate are directly observed under various water pressures. Results indicate a slight change in fracture hydraulic head leads to an observable change in aperture. This suggests that aperture change should be considered when analyzing hydraulic test data from a sparsely fractured rock aquifer. © 2012, The Author(s). Groundwater © 2012, National Ground Water Association.

  2. Determination of Material Strengths by Hydraulic Bulge Test

    Directory of Open Access Journals (Sweden)

    Hankui Wang

    2016-12-01

    Full Text Available The hydraulic bulge test (HBT method is proposed to determine material tensile strengths. The basic idea of HBT is similar to the small punch test (SPT, but inspired by the manufacturing process of rupture discs—high-pressure hydraulic oil is used instead of punch to cause specimen deformation. Compared with SPT method, the HBT method can avoid some of influence factors, such as punch dimension, punch material, and the friction between punch and specimen. A calculation procedure that is entirely based on theoretical derivation is proposed for estimate yield strength and ultimate tensile strength. Both conventional tensile tests and hydraulic bulge tests were carried out for several ferrous alloys, and the results showed that hydraulic bulge test results are reliable and accurate.

  3. Determination of Material Strengths by Hydraulic Bulge Test

    Science.gov (United States)

    Wang, Hankui; Xu, Tong; Shou, Binan

    2016-01-01

    The hydraulic bulge test (HBT) method is proposed to determine material tensile strengths. The basic idea of HBT is similar to the small punch test (SPT), but inspired by the manufacturing process of rupture discs—high-pressure hydraulic oil is used instead of punch to cause specimen deformation. Compared with SPT method, the HBT method can avoid some of influence factors, such as punch dimension, punch material, and the friction between punch and specimen. A calculation procedure that is entirely based on theoretical derivation is proposed for estimate yield strength and ultimate tensile strength. Both conventional tensile tests and hydraulic bulge tests were carried out for several ferrous alloys, and the results showed that hydraulic bulge test results are reliable and accurate. PMID:28772379

  4. THERMIT2. BWR & PWR Thermal-Hydraulic Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kazimi, M.S.; Kao, S.P.; Kelly, J.E. [Massachusetts Institute of Technology, Cambridge, MA (United States)

    1992-02-27

    THERMIT2, the most recent release of THERMIT, is intended for thermal-hydraulic analysis of both boiling and pressurized water reactor cores. It solves the three-dimensional, two-fluid equations describing the two-phase flow and heat transfer dynamics in rectangular coordinates. The two-fluid model uses separate partial differential equations expressing conservation of mass, momentum, and energy for each fluid. By expressing the exchange of mass, momentum, and energy between the fluids with physically-based mathematical models, the relative motion and thermal non-equilibrium between the fluids can exist. THERMIT2 offers the choice of either pressure or velocity boundary conditions at the top and bottom of the core. THERMIT2 includes a two-phase turbulent mixing model which provides subchannel analysis capability. THERMIT2 also solves the radial heat conduction equations for fuel pin temperatures, and calculates the heat flux from fuel pin to coolant with appropriate heat transfer models described by a boiling curve.

  5. 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

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

    Directory of Open Access Journals (Sweden)

    Yangping Zhou

    2012-01-01

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

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

    Science.gov (United States)

    Maruyama, Soh; Fujimoto, Nozomu; Kiso, Yoshihiro; Murakami, Tomoyuki; Sudo, Yukio

    1988-09-01

    This report presents the verification results of the combined thermal-hydraulic and heat conduction analysis code, FLOWNET/TRUMP which has been utilized for the core thermal hydraulic design, especially for the analysis of flow distribution among fuel block coolant channels, the determination of thermal boundary conditions for fuel block stress analysis and the estimation of fuel temperature in the case of fuel block coolant channel blockage accident in the design of the High Temperature Engineering Test Reactor(HTTR), which the Japan Atomic Energy Research Institute has been planning to construct in order to establish basic technologies for future advanced very high temperature gas-cooled reactors and to be served as an irradiation test reactor for promotion of innovative high temperature new frontier technologies. The verification of the code was done through the comparison between the analytical results and experimental results of the Helium Engineering Demonstration Loop Multi-channel Test Section(HENDEL T(sub 1-M)) with simulated fuel rods and fuel blocks.

  8. Thermal-Hydraulic Performance of the TREAT Multi-SERTTA for Reactivity Initiated Accident Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, Colby B.; Folsom, Charles P.; Davis, Cliff B.; Woolstenhulme, Nicolas E.; Bess, John D.; O' Brien, Robert C.; Ban, Heng; Wachs, Daniel M.

    2016-08-01

    Experimental testing in the Multi-Static Environment Rodlet Transient Test Apparatus (SERTTA) will lead the rebirth of transient fuel testing in the United States as part of the Accident Tolerant Fuels (ATF) progam. The Multi-SERTTA is comprised of four isolated pressurized environments capable of a wide variety of working fluids and thermal conditions. Ultimately, the TREAT reactor as well as the Multi-SERTTA test vehicle serve the purpose of providing desired thermal-hydraulic boundary conditions to the test specimen. The initial ATF testing in TREAT will focus on reactivity insertion accident (RIA) events using both gas and water environments including typical PWR operating pressures and temperatures. For the water test environment, a test configuration is envisioned using the expansion tank as part of the gas-filled expansion volume seen by the test to provide additional pressure relief. The heat transfer conditions during the high energy power pulses of RIA events remains a subject of large uncertainty and great importance for fuel performance predictions. To support transient experiments, the Multi-SERTTA vehicle has been modeled using RELAP5 with a baseline test specimen composed of UO2 fuel in zircaloy cladding. The modeling results show the influence of the designs of the specimen, vehicle, and transient power pulses. The primary purpose of this work is to provide input and boundary conditions to fuel performance code BISON. Therefore, studies of parameters having influence on specimen performance during RIA transients are presented including cladding oxidation, power pulse magnitude and width, cladding-to-coolant heat fluxes, fuel-to-cladding gap, transient boiling effects (modified CHF values), etc. The results show the great flexibility and capacity of the TREAT Multi-SERTTA test vehicle to provide testing under a wide range of prototypic thermal-hydraulic conditions as never done before.

  9. APT target/blanket design and thermal hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Cappiello, M.; Pitcher, E.; Pasamehmetoglu, K.

    1999-04-01

    The Accelerator Production of Tritium (APT) Target/Blanket (T/B) system is comprised of an assembly of tritium producing modules supported by control, heat removal, shielding and retargeting systems. The T/B assembly produces tritium using a high-energy proton beam, a tungsten/lead spallation neutron source and {sup 3}He gas as the tritium producing feedstock. For the nominal production mode, protons are accelerated to an energy of 1030 MeV at a current of 100 mA and are directed onto the T/B assembly. The protons are expanded using a raster/expansion system to illuminate a 0.19m by 1.9m beam spot on the front face of a centrally located tungsten neutron source. A surrounding lead blanket produces additional neutrons from scattered high-energy particles. The tungsten neutron source consists of nested, Inconel-718 clad tungsten cylinders assembled in horizontal Inconel-718 tubes. Each tube contains up to 6 cylinders with annular flow channel gaps of 0.102 cm. These horizontal tubes are manifolded into larger diameter vertical inlet and outlet pipes, which provide coolant. The horizontal and vertical tubes make up a structure similar to that of rungs on a ladder. The entire tungsten neutron source consists of 11 such ladders separated into two modules, one containing five ladders and the other six. Ladders are separated by a 0.3 m void region to increase nucleon leakage. The peak thermal-hydraulic conditions in the tungsten neutron source occur in the second ladder from the front. Because tungsten neutron source design has a significant number of parallel flow channels, the limiting thermal-hydraulic parameter is the onset of significant void (OSV) rather than critical heat flux (CHF). A blanket region surrounds the tungsten neutron source. The lateral blanket region is approximately 120 cm thick and 400 cm high. Blanket material consists of lead, {sup 3}He gas, aluminum, and light-water coolant. The blanket region is subdivided into rows based on the local power

  10. Laboratory tests of hydraulic fracturing and swell healing

    DEFF Research Database (Denmark)

    Thunbo, Christensen Claes; Foged, Christensen Helle; Foged, Niels

    1998-01-01

    New laboratory test set-ups and test procedures are described - for testing the formation of hydraulically induced fractures as well as the potential for subsequent fracture closurefrom the relase of a swelling potential. The main purpose with the tests is to provide information on fracturing...... stresses and whether or not the material in question possesses the potential for fracture closure....

  11. 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)

  12. 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.

  13. Hydraulic Hybrid Parcel Delivery Truck Deployment, Testing & Demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Gallo, Jean-Baptiste [Calstart Incorporated, Pasadena, CA (United States)

    2014-03-07

    Although hydraulic hybrid systems have shown promise over the last few years, commercial deployment of these systems has primarily been limited to Class 8 refuse trucks. In 2005, the Hybrid Truck Users Forum initiated the Parcel Delivery Working Group including the largest parcel delivery fleets in North America. The goal of the working group was to evaluate and accelerate commercialization of hydraulic hybrid technology for parcel delivery vehicles. FedEx Ground, Purolator and United Parcel Service (UPS) took delivery of the world’s first commercially available hydraulic hybrid parcel delivery trucks in early 2012. The vehicle chassis includes a Parker Hannifin hydraulic hybrid drive system, integrated and assembled by Freightliner Custom Chassis Corp., with a body installed by Morgan Olson. With funding from the U.S. Department of Energy, CALSTART and its project partners assessed the performance, reliability, maintainability and fleet acceptance of three pre-production Class 6 hydraulic hybrid parcel delivery vehicles using information and data from in-use data collection and on-road testing. This document reports on the deployment of these vehicles operated by FedEx Ground, Purolator and UPS. The results presented provide a comprehensive overview of the performance of commercial hydraulic hybrid vehicles in parcel delivery applications. This project also informs fleets and manufacturers on the overall performance of hydraulic hybrid vehicles, provides insights on how the technology can be both improved and more effectively used. The key findings and recommendations of this project fall into four major categories: -Performance, -Fleet deployment, -Maintenance, -Business case. Hydraulic hybrid technology is relatively new to the market, as commercial vehicles have been introduced only in the past few years in refuse and parcel delivery applications. Successful demonstration could pave the way for additional purchases of hydraulic hybrid vehicles throughout the

  14. Thermal Insulation Test Apparatuses

    Science.gov (United States)

    Berman, Brion

    2005-01-01

    The National Aeronautics and Space Administration (NASA) seeks to license its Thermal Insulation Test Apparatuses. Designed by the Cryogenics Test Laboratory at the John F. Kennedy Space Center (KSC) in Florida, these patented technologies (U.S. Patent Numbers: Cryostat 1 - 6,742,926, Cryostat 2 - 6,487,866, and Cryostat 4 - 6,824,306) allow manufacturers to fabricate and test cryogenic insulation at their production and/or laboratory facilities. These new inventions allow for the thermal performance characterization of cylindrical and flat specimens (e.g., bulk-fill, flat-panel, multilayer, or continuously rolled) over the full range of pressures, from high vacuum to no vacuum, and over the full range of temperatures from 77K to 300K. In today's world, efficient, low-maintenance, low-temperature refrigeration is taking a more significant role, from the food industry, transportation, energy, and medical applications to the Space Shuttle. Most countries (including the United States) have laws requiring commercially available insulation materials to be tested and rated by an accepted methodology. The new Cryostat methods go beyond the formal capabilities of the ASTM methods to provide testing for real systems, including full-temperature differences plus full-range vacuum conditions.

  15. IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)

    Energy Technology Data Exchange (ETDEWEB)

    Yamada, K. [Vienna International Centre, P.O. Box 100, 1400 Vienna (Austria); Aksan, S. N. [International Atomic Energy Agency, 1400 Vienna (Austria)

    2012-07-01

    The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present, 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Saha, P., E-mail: pradip.saha@ge.com [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Aksan, N. [GRNSPG Group, University of Pisa (Italy); Andersen, J. [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Yan, J. [Westinghouse Electric Co., Columbia, SC (United States); Simoneau, J.P. [AREVA, Lyon (France); Leung, L. [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada); Bertrand, F. [CEA, DEN, DER, F-13108 Saint-Paul-Lez-Durance (France); Aoto, K.; Kamide, H. [Japan Atomic Energy Agency, Chiyoda-ku, Tokyo (Japan)

    2013-11-15

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

  17. 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.

  18. 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.).

  19. 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.

  20. Test Rig Design and Presentation for a Hydraulic Yaw System

    DEFF Research Database (Denmark)

    Stubkier, Søren; Pedersen, Henrik C.; Andersen, Torben Ole

    2013-01-01

    The design and development of a hydraulic yaw system for multi MWturbines is presented and the concept explained. As part of the development of the new concept a full scale test rig for a 5 MW wind turbine has been designed and constructed. The test rig is presented along with its unique design...

  1. Periodic Hydraulic Tests in a Bedrock Fracture Network

    Science.gov (United States)

    Cole, M. C.; Becker, M.; Ciervo, C.

    2016-12-01

    Better understanding of groundwater flow through bedrock fracture networks is critical for the emerging field of enhanced geothermal systems, as well as traditional hydrogeologic characterization. Periodic hydraulic testing has shown promise for its sensitivity to local heterogeneity and, therefore, may provide useful information about flow channelization and short circuiting. Unlike conventional steady-rate pumping or injection tests, periodic tests create a disturbance such that heads in the pumping and observation wells are always in the transient state. The volume of hydraulic influence of the oscillating flow increases with period of oscillation. Thus, different portions of the formation may be interrogated even with a single well pair. We recently performed periodic pumping tests at the Mirror Lake experimental fractured rock hydrology field site in the Northeastern United States. Head in one well was oscillated while heads in five monitoring wells 30 to 60 m away were monitored. Head oscillation was accomplished through alternating injection and pumping from a surface tank and pressure was measured using a network of transducers in zones isolated by pneumatic packers. Periodicity of the induced signal was varied in order to investigate different volumes of the formation. Drawdown data from the monitoring wells were digitally filtered, which enabled use of responses that were too small or noisy for curve fitting methods. As expected, the volume of hydraulic influence increased with period, but well response was not strictly a function of distance from the source well. This anomalous response is attributed to variation in fracture network hydraulic connectivity. The ability to vary the effective penetration distance of hydraulic influence provided more information about network connectivity than from a constant rate pumping test. Estimates of hydraulic parameters displayed a decreasing trend with period length, which has been noted in previous periodic tests

  2. State of the art hydraulic turbine model test

    Science.gov (United States)

    Fabre, Violaine; Duparchy, Alexandre; Andre, Francois; Larroze, Pierre-Yves

    2016-11-01

    Model tests are essential in hydraulic turbine development and related fields. The methods and technologies used to perform these tests show constant progress and provide access to further information. In addition, due to its contractual nature, the test demand evolves continuously in terms of quantity and accuracy. Keeping in mind that the principal aim of model testing is the transposition of the model measurements to the real machine, the measurements should be performed accurately, and a critical analysis of the model test results is required to distinguish the transposable hydraulic phenomena from the test rig interactions. Although the resonances’ effects are known and described in the IEC standard, their identification is difficult. Leaning on a strong experience of model testing, we will illustrate with a few examples of how to identify the potential problems induced by the test rig. This paper contains some of our best practices to obtain the most accurate, relevant, and independent test-rig measurements.

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

    Directory of Open Access Journals (Sweden)

    M. Avramova

    2013-01-01

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

  4. Thermal energy test apparatus

    Science.gov (United States)

    Audet, N. F.

    1991-10-01

    The Navy Clothing and Textile Research Facility (NCTRF) designed and fabricated a thermal energy test apparatus to permit evaluation of the heat protection provided by crash crew firefighter's proximity clothing materials against radiant and convective heat loads, similar to those found outside the flame zone of aircraft fuel fires. The apparatus employs electrically operated quartz lamp radiant heaters and a hot air convective heater assembly to produce the heat load conditions the materials to be subjected to, and is equipped with heat flux sensors of different sensitivities to measure the incident heat flux on the sample material as well as the heat flux transmitted by the sample. Tests of the apparatus have shown that it can produce radiant heat flux levels equivalent to those estimated to be possible in close proximity to large aircraft fuel fires, and can produce convective heat fluxes equivalent to those measured in close proximity to aircraft fuel fires at upwind and sidewind locations. Work was performed in 1974.

  5. Hydraulic Model Tests on Modified Wave Dragon

    DEFF Research Database (Denmark)

    Hald, Tue; Lynggaard, Jakob

    are found in Hald and Lynggaard (2001). Model tests and reconstruction are carried out during the phase 3 project: ”Wave Dragon. Reconstruction of an existing model in scale 1:50 and sequentiel tests of changes to the model geometry and mass distribution parameters” sponsored by the Danish Energy Agency...... (DEA) wave energy programme. The tests will establish a well documented basis for the development of a 1:4.5 scale prototype planned for testing Nissum Bredning, a sea inlet on the Danish West Coast....

  6. Thermal-hydraulic and structural safety analysis of SLSF P3 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Ragland, W A; Ariman, T; Tessier, J H

    1979-01-01

    The Sodium Loop Safety Facility (SLSF) P3 experiment was the fourth in a series of in-reactor tests in the Engineering Test Reactor and simulated an unprotected flow coastdown in a 37 pin bundle. A comprehensive thermal-hydraulic analysis of the SLSF loop was coupled with a structural analysis of the test section hexcan outer duct in order to provide assurance, before the transient, that the loop would safely contain the P3 experiment. This analysis was performed for both the expected transient and for an off-normal condition of failure of redundant logic circuits to provide the proper pump power demand signal. Results of the analysis showed the hexcan outer duct to safely contain the P3 experiment for both conditions. The analysis for the expected transient has been verified by the successful completion of the P3 experiment.

  7. Development of thermal-hydraulic models for the safety evaluation of CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Young; Jung, Yun Sik; Hwang, Gi Suk; Kim, Nam Seok [Handong Univ., Pohang (Korea, Republic of); No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2004-02-15

    The objective of the present research is to evaluate the safety analysis for CANDU and to improve the Horizontal Stratification Entrainment Model (HSEM) of RELAP5/MOD3.3. This report includes two items the one is the experimental study of entrainment at horizontal pipe with {+-} 36 .deg. C , {+-} 72 .deg. C branch pies, the other is the model improvement of the moderator heat sink in the Calandria. The off-take experiments on onset of entrainment and branch quality were investigated by using water and air as working fluid, and the experimental data were compared by the previous correlations. The previous correlations could not expect experimental results, thus the weak points of the previous correlations were investigated. The improvement of the previous model continues as the next year research. The thermal hydraulic scaling analysis of SPEL, STERN and ideal linear scaling analysis have been studied. As a result, a new scaling method were needed to design a new experimental facility (HGU). A new scaling method with 1/8 length scale was applied. From these results, the thermal hydraulic model for CFD code simulation was designed and test apparatus has been made. The moderator temperature distribution experiments and CFD code simulation will be continued in next year.

  8. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, M. K.; Lee, W. J. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-03-01

    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 fourth step of the whole project, applying the RELAP5/MOD3/CANDU+ version for the real CANDU plant LOCA Analysis and D2O leakage incident. There are three main models under investigation, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs, especially when CANDU LOCA is tested. Also, for Wolsung unit 1 D2O leakage incident analysis, the plant behavior is predicted with the newly developed version for the first 1000 seconds after onset of the incident, with the main interest aiming for system pressure, level control system, and thermal hydraulic transient behavior of the secondary system. The model applied for this particular application includes heat transfer model of nuclear fuel assembly, decay heat model, and MOV (Motor Operated Valve) model. Finally, the code maintenance work, mainly correcting the known errors, is presented. 12 refs., 26 figs., 3 tabs. (Author)

  10. MODEL TESTING OF LOW PRESSURE HYDRAULIC TURBINE WITH HIGHER EFFICIENCY

    Directory of Open Access Journals (Sweden)

    V. K. Nedbalsky

    2007-01-01

    Full Text Available A design of low pressure turbine has been developed and it is covered by an invention patent and a useful model patent. Testing of the hydraulic turbine model has been carried out when it was installed on a vertical shaft. The efficiency was equal to 76–78 % that exceeds efficiency of the known low pressure blade turbines. 

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

    Science.gov (United States)

    Alkaabi, Ahmed K.

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

  12. Validation of CESAR Thermal-hydraulic Module of ASTEC V1.2 Code on BETHSY Experiments

    Science.gov (United States)

    Tregoures, Nicolas; Bandini, Giacomino; Foucher, Laurent; Fleurot, Joëlle; Meloni, Paride

    The ASTEC V1 system code is being jointly developed by the French Institut de Radioprotection et Sûreté Nucléaire (IRSN) and the German Gesellschaft für Anlagen und ReaktorSicherheit (GRS) to address severe accident sequences in a nuclear power plant. Thermal-hydraulics in primary and secondary system is addressed by the CESAR module. The aim of this paper is to present the validation of the CESAR module, from the ASTEC V1.2 version, on the basis of well instrumented and qualified integral experiments carried out in the BETHSY facility (CEA, France), which simulates a French 900 MWe PWR reactor. Three tests have been thoroughly investigated with CESAR: the loss of coolant 9.1b test (OECD ISP N° 27), the loss of feedwater 5.2e test, and the multiple steam generator tube rupture 4.3b test. In the present paper, the results of the code for the three analyzed tests are presented in comparison with the experimental data. The thermal-hydraulic behavior of the BETHSY facility during the transient phase is well reproduced by CESAR: the occurrence of major events and the time evolution of main thermal-hydraulic parameters of both primary and secondary circuits are well predicted.

  13. 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.

  14. Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C., E-mail: sabrinapral@gmail.com, E-mail: amir@cdtn.brm, E-mail: hcr@cdtn.br, E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

  15. History of the 185-/189-D thermal hydraulics laboratory and its effects on reactor operations at the Hanford Site

    Energy Technology Data Exchange (ETDEWEB)

    Gerber, M.S.

    1994-09-01

    The 185-D deaeration building and the 189-D refrigeration building were constructed at Hanford during 1943 and 1944. Both buildings were constructed as part of the influent water cooling system for D reactor. The CMS studies eliminated the need for 185-D function. Early gains in knowledge ended the original function of the 189-D building mission. In 1951, 185-D and 189-D were converted to a thermal-hydraulic laboratory. The experiments held in the thermal-hydraulic lab lead to historic changes in Hanford reactor operations. In late 1951, the exponential physics experiments were moved to the 189-D building. In 1958, new production reactor experiments were begun in 185/189-D. In 1959, Plutonium Recycle Test Reactor experiments were added to the 185/189-D facility. By 1960, the 185/189-D thermal hydraulics laboratory was one of the few full service facilities of its type in the nation. During the years 1961--1963 tests continued in the facility in support of existing reactors, new production reactors, and the Plutonium Recycle Test Reactor. In 1969, Fast Flux Test Facility developmental testings began in the facility. Simulations in 185/189-D building aided in the N Reactor repairs in the 1980`s. In 1994 the facility was nominated to the National Register of Historic Places, because of its pioneering role over many years in thermal hydraulics, flow studies, heat transfer, and other reactor coolant support work. During 1994 and 1995 it was demolished in the largest decontamination and decommissioning project thus far in Hanford Site history.

  16. 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)

  17. Analysis of Slug Tests in Formations of High Hydraulic Conductivity

    Science.gov (United States)

    Butler, J.J.; Garnett, E.J.; Healey, J.M.

    2003-01-01

    A new procedure is presented for the analysis of slug tests performed in partially penetrating wells in formations of high hydraulic conductivity. This approach is a simple, spreadsheet-based implementation of existing models that can be used for analysis of tests from confined or unconfined aquifers. Field examples of tests exhibiting oscillatory and nonoscillatory behavior are used to illustrate the procedure and to compare results with estimates obtained using alternative approaches. The procedure is considerably simpler than recently proposed methods for this hydrogeologic setting. Although the simplifications required by the approach can introduce error into hydraulic-conductivity estimates, this additional error becomes negligible when appropriate measures are taken in the field. These measures are summarized in a set of practical field guidelines for slug tests in highly permeable aquifers.

  18. Origin of honeycombs: Testing the hydraulic and case hardening hypotheses

    Science.gov (United States)

    Bruthans, Jiří; Filippi, Michal; Slavík, Martin; Svobodová, Eliška

    2018-02-01

    Cavernous weathering (cavernous rock decay) is a global phenomenon, which occurs in porous rocks around the world. Although honeycombs and tafoni are considered to be the most common products of this complex process, their origin and evolution are as yet not fully understood. The two commonly assumed formation hypotheses - hydraulic and case hardening - were tested to elucidate the origin of honeycombs on sandstone outcrops in a humid climate. Mechanical and hydraulic properties of the lips (walls between adjacent pits) and backwalls (bottoms of pits) of the honeycombs were determined via a set of established and novel approaches. While the case hardening hypothesis was not supported by the determinations of either tensile strength, drilling resistance or porosity, the hydraulic hypothesis was clearly supported by field measurements and laboratory tests. Fluorescein dye visualization of capillary zone, vapor zone, and evaporation front upon their contact, demonstrated that the evaporation front reaches the honeycomb backwalls under low water flow rate, while the honeycomb lips remain dry. During occasional excessive water flow events, however, the evaporation front may shift to the lips, while the backwalls become moist as a part of the capillary zone. As the zone of evaporation corresponds to the zone of potential salt weathering, it is the spatial distribution of the capillary and vapor zones which dictates whether honeycombs are created or the rock surface is smoothed. A hierarchical model of factors related to the hydraulic field was introduced to obtain better insights into the process of cavernous weathering.

  19. 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.

  20. Creation of an Enhanced Geothermal System through Hydraulic and Thermal Stimulation

    Energy Technology Data Exchange (ETDEWEB)

    Rose, Peter Eugene [Energy and Geoscience Institute at the University of Utah

    2013-04-15

    This report describes a 10-year DOE-funded project to design, characterize and create an Engineered Geothermal System (EGS) through a combination of hydraulic, thermal and chemical stimulation techniques. Volume 1 describes a four-year Phase 1 campaign, which focused on the east compartment of the Coso geothermal field. It includes a description of the geomechanical, geophysical, hydraulic, and geochemical studies that were conducted to characterize the reservoir in anticipation of the hydraulic stimulation experiment. Phase 1 ended prematurely when the drill bit intersected a very permeable fault zone during the redrilling of target stimulation well 34-9RD2. A hydraulic stimulation was inadvertently achieved, however, since the flow of drill mud from the well into the formation created an earthquake swarm near the wellbore that was recorded, located, analyzed and interpreted by project seismologists. Upon completion of Phase 1, the project shifted focus to a new target well, which was located within the southwest compartment of the Coso geothermal field. Volume 2 describes the Phase 2 studies on the geomechanical, geophysical, hydraulic, and geochemical aspects of the reservoir in and around target-stimulation well 46A-19RD, which is the deepest and hottest well ever drilled at Coso. Its total measured depth exceeding 12,000 ft. It spite of its great depth, this well is largely impermeable below a depth of about 9,000 ft, thus providing an excellent target for stimulation. In order to prepare 46A-19RD for stimulation, however, it was necessary to pull the slotted liner. This proved to be unachievable under the budget allocated by the Coso Operating Company partners, and this aspect of the project was abandoned, ending the program at Coso. The program then shifted to the EGS project at Desert Peak, which had a goal similar to the one at Coso of creating an EGS on the periphery of an existing geothermal reservoir. Volume 3 describes the activities that the Coso team

  1. Hydraulic testing of simulated DWPF waste slurries at the Georgia Iron Works Hydraulic Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, D.P.

    1982-12-31

    Pipeline tests of current simulations of Defense Waste Processing Facility (DWPF) waste slurries were performed during August 1982 at the Georgia Iron Works Hydraulic Laboratory (GIW). Measurements of pressure gradient versus flow in 3-inch pipes and fittings were made for various concentrations of unformated sludge and formated sludge with frit. All slurries were shown to behave generally as Bingham Plastic fluids. Formated sludge/frit slurries behaved generally like unformated sludge slurries of comparable yield stress. No frit settling problems were observed. 8 refs., 16 figs., 6 tabs.

  2. Improving Neutron Kinetics and Thermal Hydraulics coupled tools for BEPU calculations

    Energy Technology Data Exchange (ETDEWEB)

    Pericas, R.; Reventós, F.; Batet, Il.

    2015-07-01

    The BEPU methodology is capable of providing a solution in terms of increasing the nuclear power production without compromising the safety margins. This study presents different improvements performed using tools available at UPC in the field of Neutron Kinetics and Thermal Hydraulics coupled systems. The paper describes a comparison between the BEPU methodology and the Conservative Bounding methodology within the framework of the Neutron Kinetics and Thermal Hydraulics coupled systems. To perform such comparison the following tools have been selected: TRACE for thermal-hydraulic system calculations, PARCS for reactor kinetics core simulator code. A Main Steam Line Break (MSLB) in a Pressurized Water Reactor (PWR) is the selected simulated transient to show the improvements performed. (Author)

  3. Flammability Tests on Hot Surface for Several Hydraulic Fluids

    Directory of Open Access Journals (Sweden)

    L. Deleanu

    2011-09-01

    Full Text Available Industrial equipment using hydraulic fluids are design to accept higher load and speed, implicitly higher temperatures, including for fluids. Leakages from enclosures like gear boxes or hydraulic systems could increase the risk of fluid reaching hot surfaces, thus producing fires hard to be controlled and isolated. The designer have to evaluate the flammability of fluids and they should select several solutions for a particular application in order to estimate the costs of different solutions and to mitigate the risk of having accidental fires due to a specific fluid grade. The tests were done with the help of an original equipment allowing a dedicated soft assistance in order to protect the operator and to sustain reproducibility, according to the standard SR EN ISO 20823:2004 Petroleum and related products. The determination of the flammability characteristics of fluids in contact with hot surfaces - Manifold ignition test, There were tested the following grades of hydraulic oil HLP 68 X-Oil, HFC Prista, MHE 40 Prista (100% oil, a rapeseed oil (obtained after a dewaxing process and an emulsion oil-in-water (5% vol. MHE 40 Prista. There were identified distinct behaviours of these fluids under the test conditions

  4. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  6. Development of thermal hydraulic models for main circulation circuit of RBMK-1500 reactor using Apros and Cathare 2 codes

    Energy Technology Data Exchange (ETDEWEB)

    Zemulis, G.; Jasiulevicius, A. [Kaunas University of Technology, Dept. of Thermal and Nuclear Energy, Kaunas, (Lithuania)

    2001-07-01

    Reactor safety is the most important issue in nuclear engineering. It concerns the capability of the nuclear object to withhold the main safety and reliability criterion within specified range during both normal operation and transient conditions. Three types of assessment are to be performed in order to establish the nuclear power plant safety level: neutronic calculations; thermal hydraulic calculations; mechanical design calculations. Calculations of the thermal hydraulic parameters of the RBMK-1500 reactor main circulation circuit (MCC) are presented in this paper. The aim of this work was to test the capability of the APROS code to simulate the behavior of the RBMK-1500 type reactor main circulation circuit during normal operation and transients. (author)

  7. Workgroup for Hydraulic laboratory Testing and Verification of Hydroacoustic Instrumentation

    Science.gov (United States)

    Fulford, Janice M.; Armstrong, Brandy N.; Thibodeaux, Kirk G.

    2015-01-01

    An international workgroup was recently formed for hydraulic laboratory testing and verification of hydroacoustic instrumentation used for water velocity measurements. The activities of the workgroup have included one face to face meeting, conference calls and an inter-laboratory exchange of two acoustic meters among participating laboratories. Good agreement was found among four laboratories at higher tow speeds and poorer agreement at the lowest tow speed.

  8. Thermal and Hydraulic Performances of Nanofluids Flow in Microchannel Heat Sink with Multiple Zigzag Flow Channels

    Directory of Open Access Journals (Sweden)

    Duangthongsuk Weerapun

    2017-01-01

    Full Text Available This article presents an experimental investigation on the heat transfer performance and pressure drop characteristic of two types of nanofluids flowing through microchannel heat sink with multiple zigzag flow channel structures (MZMCHS. SiO2 nanoparticles dispersed in DI water with concentrations of 0.3 and 0.6 vol.% were used as working fluid. MZMCHS made from copper material with dimension of 28 × 33 mm. Hydraulic diameter of MZMCHs is designed at 1 mm, 7 number of flow channels and heat transfer area is about 1,238 mm2. Effects of particle concentration and flow rate on the thermal and hydraulic performances are determined and then compare with the common base fluid. The results indicated that the heat transfer coefficient of nanofluids was higher than that of the water and increased with increasing particle concentration as well as Reynolds number. For pressure drop, the particle concentrations have no significant effect on the pressure drop across the test section.

  9. Characterization of thermal, hydraulic, and gas diffusion properties in variably saturated sand grades

    DEFF Research Database (Denmark)

    Deepagoda Thuduwe Kankanamge Kelum, Chamindu; Smits, Kathleen; Ramirez, Jamie

    2016-01-01

    porous media transport properties, key transport parameters such as thermal conductivity and gas diffusivity are particularly important to describe temperature-induced heat transport and diffusion-controlled gas transport processes, respectively. Despite many experimental and numerical studies focusing....../70) in relation to physical properties, water retention, hydraulic conductivity, thermal conductivity, and gas diffusivity. We used measured basic properties and transport data to accurately parameterize the characteristic functions (particle- and pore-size distributions and water retention) and descriptive...... transport models (thermal conductivity, saturated hydraulic conductivity, and gas diffusivity). An existing thermal conductivity model was improved to describe the distinct three-region behavior in observed thermal conductivity–water saturation relations. Applying widely used parametric models for saturated...

  10. 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.

  11. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, J.G.; Herring, J.S.; Carlson, K.E.; Ransom, V.H.

    1981-01-01

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

  13. 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.

  14. Application of flow network models of SINDA/FLUINT{sup TM} to a nuclear power plant system thermal hydraulic code

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Ji Bum [Institute for Advanced Engineering, Yongin (Korea, Republic of); Park, Jong Woon [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    In order to enhance the dynamic and interactive simulation capability of a system thermal hydraulic code for nuclear power plant, applicability of flow network models in SINDA/FLUINT{sup TM} has been tested by modeling feedwater system and coupling to DSNP which is one of a system thermal hydraulic simulation code for a pressurized heavy water reactor. The feedwater system is selected since it is one of the most important balance of plant systems with a potential to greatly affect the behavior of nuclear steam supply system. The flow network model of this feedwater system consists of condenser, condensate pumps, low and high pressure heaters, deaerator, feedwater pumps, and control valves. This complicated flow network is modeled and coupled to DSNP and it is tested for several normal and abnormal transient conditions such turbine load maneuvering, turbine trip, and loss of class IV power. The results show reasonable behavior of the coupled code and also gives a good dynamic and interactive simulation capabilities for the several mild transient conditions. It has been found that coupling system thermal hydraulic code with a flow network code is a proper way of upgrading simulation capability of DSNP to mature nuclear plant analyzer (NPA). 5 refs., 10 figs. (Author)

  15. National Solar Thermal Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The National Solar Thermal Test Facility (NSTTF) is the only test facility in the United States of its type. This unique facility provides experimental engineering...

  16. Hydraulic tests of emergency cooling system: L-Area

    Energy Technology Data Exchange (ETDEWEB)

    Hinton, J H

    1988-01-01

    The delay in L-Area startup provided an opportunity to obtain valuable data on the Emergency Cooling System (ECS) which will permit reactor operation at the highest safe power level. ECS flow is a major input to the FLOOD code which calculates reactor ECS power limits. The FLOOD code assesses the effectiveness of the ECS cooling capacity by modeling the core and plenum hydraulics under accident conditions. Presently, reactor power is not limited by the ECS cooling capacity (power limit). However, the manual calculations of ECS flows had been recently updated to include piping changes (debris strainer, valve changes, pressure release systems) and update fitting losses. Both updates resulted in reduced calculated ECS flows. Upon completion of the current program to update, validate, and document, reactor power may be limited under certain situations by ECS cooling capacity for some present reactor charge designs. A series of special hydraulic tests (Reference 1, 3) were conducted in L-Area using all sources of emergency coolant including the ECS pumps (Reference 2). The tests provided empirical hydraulic data on the ECS piping. These data will be used in computer models of the system as well as manual calculations of ECS flows. The improved modeling and accuracy of the flow calculations will permit reactor operation at the highest safe power level with respect to an ECS power limit.

  17. 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)

  18. 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.

  19. Hydraulic assessment of the Buda Thermal Karst area and its vulnerability (Budapest, Hungary)

    Science.gov (United States)

    Czauner, Brigitta; Erőss, Anita; Erhardt, Ildikó; Ötvös, Viktória; Simon, Szilvia; Mádl-Szőnyi, Judit

    2017-04-01

    Thermal and medicinal water resources of Budapest (Hungary), the "City of Spas", are provided by the Buda Thermal Karst area. Assessment of its vulnerability requires the understanding of the discharge phenomena and thus the groundwater flow conditions in the area. Accordingly, BTK has already been the objective of several hydrogeological investigations, including numerical simulations as well, which led to conceptual models. The aim of the present study was the hydraulic evaluation of the flow systems based on the complex analysis of real, i.e. measured, archival hydraulic data of wells in order to i) get acquainted with the real flow systems, and ii) hydraulically confirm or disprove the previous conceptual models, in particular the applicability of gravity-driven regional groundwater flow concept and hydraulic continuity, separation of the natural discharge zones, and hypogenic karstification. Considering the data distribution, pressure vs. elevation profiles, tomographic fluid-potential maps, and hydraulic cross-sections were constructed for the first time in this area. As a result, gravitational flow systems and the modifying effects of aquitard units and faults were identified. Consequently, the differences in temperature, hydrochemistry, discharge distribution (one and two-components), and related cave forming processes between the Central (Rózsadomb) and Southern (Gellért Hill) natural discharge areas could be explained, as well as the hydraulic behaviour of the Northeastern Margin-fault of the Buda Hills could be determined. Regarding the on-going hypogenic karstification processes, regional upward flow conditions were confirmed along the main discharge zone of the Danube. Identification of gravity as the main fluid flow driving force, as well as the hydraulic effects of heterogeneities can significantly contribute to the recognition of the risk factors regarding the vulnerability of the Buda Thermal Karst area. The research was supported by the

  20. The time-dependent 3D discrete ordinates code TORT-TD with thermal-hydraulic feedback by ATHLET models

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

    This paper describes the time-dependent 3D discrete ordinates transport code TORT-TD. Thermal-hydraulic feedback is considered by coupling TORT-TD with the thermal-hydraulics system code ATHLET. The coupled code TORT-TD/ATHLET allows 3D pin-by-pin analyses of transients in few energy groups and anisotropic scattering by solving the time-dependent transport equation using the unconditionally stable implicit method. The nuclear cross sections are interpolated between pre-calculated table values of fuel temperature, moderator density and boron concentration. For verification of the implementation, selected test cases have been calculated by TORT-TD/ATHLET. They include a control rod ejection transient in a small PWR fuel assembly arrangement and a local boron concentration change in a single PWR fuel assembly. In the latter, special attention has been paid to study the influence of the thermal-hydraulic feedback modelling in ATHLET. The results obtained for a control rod ejection accident in a PWR quarter core demonstrate the applicability of TORT-TD/ATHLET. (authors)

  1. 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.

  2. Experimental studies into the thermal-hydraulic performance of the VK-300 reactor based on a draft tube model

    Directory of Open Access Journals (Sweden)

    N.P. Serdun

    2015-12-01

    Full Text Available The paper presents an experimental study into the thermal-hydraulic performance of the VK-300 reactor based on a model of a single draft tube at a pressure of 3.4MPa, various flow rates and the model inlet relative enthalpies of –0.05 to 0.2. The experimental procedures include generation of a steam-water mixture circulation with a preset flow rate and a relative enthalpy through the test section at a pressure of 3.3 to 3.4MPa, and measurement of thermal-hydraulic parameters within the circuit's representative upflow and downflow lengths of practical interest. There have been confirmed the designs used to support the reactor facility serviceability and the assumptions concerning the thermal-hydraulic performance of a natural circulation circuit used in the analysis thereof. It has been shown that, across the analyzed range of the relative enthalpy values, the draft tube has an annular-dispersed or an annular flow of the steam-water mixture, both providing for the significant separation of the steam-water mixture (Ksep=0.4 at the draft tube edges and in the mixing chamber. The perforation in the upper part of the draft tubes allows the separation coefficient to be increased at the first stage and creates more favorable conditions for the second-stage separation. The measured values of the void fraction in the mixing chamber and in the draft tube are in a satisfactory agreement with calculations based on Z.L. Miropolskiy's method and the RELAP code and may be used to verify the VK-300 thermal-hydraulic codes. It has been shown that steam may enter the ring slit that simulates the annular space and reach the reactor core inlet. Further investigations need to be conducted to study this effect for its guaranteed exclusion and for the development of emergency response procedures.

  3. Analytical Thermal Field Theory Applicable to Oil Hydraulic Fluid Film Lubrication

    DEFF Research Database (Denmark)

    Johansen, Per; Roemer, Daniel Beck; Pedersen, Henrik C.

    2014-01-01

    An analytical thermal field theory is derived by a perturbation series expansion solution to the energy conservation equation. The theory is valid for small values of the Brinkman number and the modified Peclet number. This condition is sufficiently satisfied for hydraulic oils, whereby...

  4. Thermal hydraulic studies of spallation target for one-way coupled ...

    Indian Academy of Sciences (India)

    pp. 355–363. Thermal hydraulic studies of spallation target for one-way coupled Indian accelerator driven systems with low energy proton beam. V MANTHA1, A K MOHANTY2 and P SATYAMURTHY1. 1Laser and Plasma Technology Division; 2Nuclear Physics Division, Bhabha Atomic. Research Centre, Mumbai 400 085, ...

  5. 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.

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

    Indian Academy of Sciences (India)

    Home; Journals; Sadhana; Volume 40; Issue 3. A review on the thermal hydraulic characteristics of the air-cooled heat exchangers in forced convection. Ankur Kumar Jyeshtharaj B Joshi Arun K Nayak Pallippattu K Vijayan. Section I – Fluid Mechanics and Fluid Power (FMFP) Volume 40 Issue 3 May 2015 pp 673-755 ...

  7. Advances in thermal hydraulic and neutronic simulation for reactor analysis and safety

    Energy Technology Data Exchange (ETDEWEB)

    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.

    1993-03-01

    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.

  8. Thermal-hydraulically corrected neutron cross-sections for PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Santiago, Daniela M.N.; Alvim, Antonio C.M.; Silva, Fernando C., E-mail: dsantiago@con.ufrj.b, E-mail: alvim@con.ufrj.b, E-mail: fernando@con.ufrj.b [Universidade Federal do Rio de Janeiro (PEN/COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear

    2011-07-01

    Reactor core simulation codes ought to have a thermal-hydraulics feedback module. This module calculates, among other effects, the fuel temperature thermal-hydraulics feedback, that corrects neutron cross sections. In the nodal code developed at PEN/COPPE/UFRJ, the fuel temperature is obtained through an empirical model. This work presents a physical model to calculate this temperature. A finite volume technique was used to discretize the equation for temperature distribution, while the moderator coefficient of heat transfer was calculated using ASME routines, appended to the developed code. This model allows calculation of an average radial temperature per node, since the thermal-hydraulic feedback must follow the conditions imposed by the nodal code. The results obtained were compared with the ones obtained by the empirical model. The results show that, for fuel elements near core 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. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Waata, C.L.

    2006-07-15

    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

  10. 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

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

    Directory of Open Access Journals (Sweden)

    Królikowski Igor P.

    2015-09-01

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

  12. Contribution of thermal–hydraulic validation tests to the standard design approval of SMART

    Directory of Open Access Journals (Sweden)

    Hyun-Sik Park

    2017-10-01

    Full Text Available Many thermal–hydraulic tests have been conducted at the Korea Atomic Energy Research Institute for verification of the SMART (System-integrated Modular Advanced ReacTor design, the standard design approval of which was issued by the Korean regulatory body. In this paper, the contributions of these tests to the standard design approval of SMART are discussed. First, an integral effect test facility named VISTA-ITL (Experimental Verification by Integral Simulation of Transients and Accidents-Integral Test Loop has been utilized to assess the TASS/SMR-S (Transient and Set-point Simulation/Small and Medium safety analysis code and confirm its conservatism, to support standard design approval, and to construct a database for the SMART design optimization. In addition, many separate effect tests have been performed. The reactor internal flow test has been conducted using the SCOP (SMART COre flow distribution and Pressure drop test facility to evaluate the reactor internal flow and pressure distributions. An ECC (Emergency Core Coolant performance test has been carried out using the SWAT (SMART ECC Water Asymmetric Two-phase choking test facility to evaluate the safety injection performance and to validate the thermal–hydraulic model used in the safety analysis code. The Freon CHF (Critical Heat Flux test has been performed using the FTHEL (Freon Thermal Hydraulic Experimental Loop facility to construct a database from the 5 × 5 rod bundle Freon CHF tests and to evaluate the DNBR (Departure from Nucleate Boiling Ratio model in the safety analysis and core design codes. These test results were used for standard design approval of SMART to verify its design bases, design tools, and analysis methodology.

  13. 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.

  14. Thermal hydraulic behavior and efficiency analysis of an all-vanadium redox flow battery

    Science.gov (United States)

    Xiong, Binyu; Zhao, Jiyun; Tseng, K. J.; Skyllas-Kazacos, Maria; Lim, Tuti Mariana; Zhang, Yu

    2013-11-01

    Vanadium redox flow batteries (VRBs) are very competitive for large-capacity energy storage in power grids and in smart buildings due to low maintenance costs, high design flexibility, and long cycle life. Thermal hydraulic modeling of VRB energy storage systems is an important issue and temperature has remarkable impacts on the battery efficiency, the lifetime of material and the stability of the electrolytes. In this paper, a lumped model including auxiliary pump effect is developed to investigate the VRB temperature responses under different operating and surrounding environmental conditions. The impact of electrolyte flow rate and temperature on the battery electrical characteristics and efficiencies are also investigated. A one kilowatt VRB system is selected to conduct numerical simulations. The thermal hydraulic model is benchmarked with experimental data and good agreement is found. Simulation results show that pump power is sensitive to hydraulic design and flow rates. The temperature in the stack and tanks rises up about 10 °C under normal operating conditions for the stack design and electrolyte volume selected. An optimal flow rate of around 90 cm3 s-1 is obtained for the proposed battery configuration to maximize battery efficiency. The models developed in this paper can also be used for the development of a battery control strategy to achieve satisfactory thermal hydraulic performance and maximize energy efficiency.

  15. Development of MCATHAS system of coupled neutronics/thermal-hydraulics in supercritical water reactor

    Energy Technology Data Exchange (ETDEWEB)

    An, P.; Yao, D. [Science and Tech. on Reactor System Design Tech. Laboratory, Chengdu (China)

    2011-07-01

    The MCATHAS system of coupled neutronics/Thermal-hydraulics in supercritical water reactor is described, which considers the mutual influence between the obvious axial and radial evolution of material temperature, water density and the relative power distribution. This system can obtain the main neutronics and thermal parameters along with burn-up. MCATHAS system is parallel processing coupling. The MCNP code is used for neutronics analysis with the continuous cross section library at any temperature calculated by interpolation algorithm; The sub-channel code ATHAS is for thermal-hydraulics analysis and the ORIGEN Code for burn-up calculation. We validate the code with the assembly of HPLWR and analyze the assembly SCLWR- H. (author)

  16. Thermal-hydraulic design of tungsten rod bundles for the APT 3He neutron spallation target

    Science.gov (United States)

    Willcutt, Gordon J. E.

    1995-01-01

    A preconceptual design has been developed for the 3He Target/Blanket System for the Accelerator Production of Tritium Project. The design use tungsten wire-wrapped rods to produce neutrons when the rods are struck by a proton beam. The rods are contained in bundles inside hexagonal Inconel ducts and cooled by D2O. Rod bundles are grouped in patterns in the proton beam inside a chamber filled with 3He that is transmuted to tritium by the neutrons coming from the tungsten rods. Additional 3He is transmuted in a blanket region surrounding the helium chamber. This paper describes the initial thermal-hydraulic design and testing that has been completed to confirm the designed calculations for pressure drop through the bundle and heat transfer in the bundle. Heat transfer tests were run to verify steady-state operation. These tests were followed by increasing power until nucleate boiling occurs to determine operating margins. Changes that improve the initial design are described.

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

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, J.; Herrero, J. J.; Cuervo, D.; Aragones, J. M., E-mail: jimenez@din.upm.e [Universidad Politecnica de Madrid, Departamento de Ingenieria Nuclear, Jose Gutierrez Abascal No. 2, 28006 Madrid (Spain)

    2010-10-15

    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)

  18. Design and Construction of Experiment for Direct Electron Irradiation of Uranyl Sulfate Solution: Bubble Formation and Thermal Hydraulics Studies

    Energy Technology Data Exchange (ETDEWEB)

    Chemerisov, Sergey [Argonne National Lab. (ANL), Argonne, IL (United States); Gromov, Roman [Argonne National Lab. (ANL), Argonne, IL (United States); Makarashvili, Vakho [Argonne National Lab. (ANL), Argonne, IL (United States); Heltemes, Thad [Argonne National Lab. (ANL), Argonne, IL (United States); Sun, Zaijing [Argonne National Lab. (ANL), Argonne, IL (United States); Wardle, Kent E. [Argonne National Lab. (ANL), Argonne, IL (United States); Bailey, James [Argonne National Lab. (ANL), Argonne, IL (United States); Quigley, Kevin [Argonne National Lab. (ANL), Argonne, IL (United States); Stepinski, Dominique [Argonne National Lab. (ANL), Argonne, IL (United States); Vandegrift, George [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-10-01

    Argonne is assisting SHINE Medical Technologies in developing SHINE, a system for producing fission-product 99Mo using a D/T-accelerator to produce fission in a non-critical target solution of aqueous uranyl sulfate. We have developed an experimental setup for studying thermal-hydraulics and bubble formation in the uranyl sulfate solution to simulate conditions expected in the SHINE target solution during irradiation. A direct electron beam from the linac accelerator will be used to irradiate a 20 L solution (sector of the solution vessel). Because the solution will undergo radiolytic decomposition, we will be able to study bubble formation and dynamics and effects of convection and temperature on bubble behavior. These experiments will serve as a verification/ validation tool for the thermal-hydraulic model. Utilization of the direct electron beam for irradiation allows homogeneous heating of a large solution volume and simplifies observation of the bubble dynamics simultaneously with thermal-hydraulic data collection, which will complement data collected during operation of the miniSHINE experiment. Irradiation will be conducted using a 30-40 MeV electron beam from the high-power linac accelerator. The total electron-beam power will be 20 kW, which will yield a power density on the order of 1 kW/L. The solution volume will be cooled on the front and back surfaces and central tube to mimic the geometry of the proposed SHINE solution vessel. Also, multiple thermocouples will be inserted into the solution vessel to map thermal profiles. The experimental design is now complete, and installation and testing are in progress.

  19. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, F.; Hibiki, T.; Saito, Y.; Takeda, T.; Mishima, K. [Kyoto Univ., Research Reactor Institute (Japan)

    2001-07-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Thomas Michael; Shadid, John N; Pawlowski, Roger P; Cyr, Eric C; Wildey, Timothy Michael

    2014-01-01

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

  2. Review of the nuclear reactor thermal hydraulic research in ocean motions

    Energy Technology Data Exchange (ETDEWEB)

    Yan, B.H., E-mail: yanbh3@mail.sysu.edu.cn

    2017-03-15

    The research and development of small modular reactor in floating platform has been strongly supported by Chinese government and enterprises. Due to the effect of ocean waves, the thermal hydraulic behavior and safety characteristics of floating reactor are different from that of land-based reactor. Many scholars including the author have published their research and results in open literatures. Much of these literatures are valuable but there are also some contradictory conclusions. In this wok, the nuclear reactor thermal hydraulic research in ocean motions was systematically summarized. Valuable results and experimental data were analyzed and classified. Inherent mechanism for controversial issues in different experiments was explained. Necessary work needed in the future was suggested. Through this work, we attempt to find as many valuable results as possible for the designing and subsequent research.

  3. Current and anticipated uses of thermal-hydraulic codes in Germany

    Energy Technology Data Exchange (ETDEWEB)

    Teschendorff, V.; Sommer, F.; Depisch, F.

    1997-07-01

    In Germany, one third of the electrical power is generated by nuclear plants. ATHLET and S-RELAP5 are successfully applied for safety analyses of the existing PWR and BWR reactors and possible future reactors, e.g. EPR. Continuous development and assessment of thermal-hydraulic codes are necessary in order to meet present and future needs of licensing organizations, utilities, and vendors. Desired improvements include thermal-hydraulic models, multi-dimensional simulation, computational speed, interfaces to coupled codes, and code architecture. Real-time capability will be essential for application in full-scope simulators. Comprehensive code validation and quantification of uncertainties are prerequisites for future best-estimate analyses.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-04-01

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

  5. The SESAME project. State of the art liquid metal thermal hydraulics and beyond

    Energy Technology Data Exchange (ETDEWEB)

    Roelofs, F.; Shams, A.; Batta, A.; Moreau, V.; Di Piazza, I.; Gerschenfeld, A.; Planquart, P.; Tarantino, M. [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands)

    2017-08-15

    The European Sustainable Nuclear Industry Initiative (ESNII) aims at industrial application of fast reactor technology for a sustainable nuclear energy production. Currently four demonstration projects have a promising outlook in Europe, i.e. the ASTRID project in France, the MYRRHA project in Belgium, the ALFRED pan-European project to be realized in Romania, and SEALER 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 discusses the state-of-the-art knowledge with respect to experiments and simulation techniques as pursued in the Horizon 2020 SESAME (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors) project.

  6. ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) solutions to developmental assessment problems

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. 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.

  8. Reactor thermal-hydraulic FY 1986 status report for the multimegawatt Space Nuclear Power Program

    Energy Technology Data Exchange (ETDEWEB)

    Krotiuk, W.J.; Antoniak, Z.I.

    1986-10-01

    PNL's 1986 activities can be divided into three basic areas: code assessment, correlation assessment and experimental activities. The ultimate goal of all these activities is developing computer codes and verifying their use to perform the thermal-hydraulic analysis and design of the reactor core and plenum of the various proposed concepts. To perform this task as assessment is made of existing computer codes, models, correlations, and microgravity experimental data.

  9. 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)

  10. 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)

  11. 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

  12. Changes of soil thermal and hydraulic regimes in the Heihe River Basin.

    Science.gov (United States)

    Peng, Xiaoqing; Mu, Cuicui

    2017-09-02

    Soil thermal and hydraulic regimes are critical factors influencing terrestrial processes in cold regions. Collection of field data from frozen ground has occurred at point scales, but limited data exist that characterize changes of soil thermal and hydraulic regimes at the scale of the whole Heihe River Basin. This study uses a long-term regional climate model coupled with land surface model to investigate the soil thermal and hydraulic regime changes at a large spatial scale. It also explores potential factors, including the climate and non-climate factors. Results show that there is significant variability in mean annual air temperature (MAAT) of about 0.47 °C/decade during 1980-2013. A time series of area-averaged mean annual soil temperature (MAST) over the whole Heihe River Basin shows a significant increase between 0.25 and 0.36 °C/decade during 1984-2013, with a net change of 0.9 °C. A trend of increasing wetness is found in soil moisture. Frozen days (FD) decreased significantly both in seasonally frozen ground (SFG) regions and permafrost regions, with a net change between 7 and 13 days during 1984-2013. Freezing index (FI) had a positive effect on FD, while thawing index (TI), MAAT, precipitation, and normalized difference vegetation index (NDVI) had a negative effect. These results are important to understand dynamic mechanisms of soil freeze/thaw cycles.

  13. Hydrogen Behaviors Coupled with Thermal Hydraulics in APR1400 Containment during an SBLOCA

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jongtae; Hong, Seong-Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    There are two hydrogen mitigation strategies. One is burning hydrogen by igniters as it is released in the containment. The other is using PARs (passive auto-catalytic recombiners) to remove the hydrogen. The strategy based on PAR installation requires well-mixing of the released hydrogen with steam in a containment atmosphere. It means that hydrogen distribution in a containment is very important for the hydrogen mitigation by PAR. Hydrogen behaviors in a NPP containment is strongly coupled with thermal hydraulics such as turbulent mixing, stratification by buoyancy, steam condensation, heat transfer et al. The purpose of this study is to investigate hydrogen behaviors in the APR1400 containment during an SBLOCA with a reactor core damage. The GASFLOW code is used to simulate hydrogen behaviors with thermal hydraulics in the APR1400 containment. In this paper, the hydrogen behaviors in the APR1400 containment during an SBLOCA were investigated by numerical simulation using GASFLOW. It was found that hydrogen mixture cloud may move downward relying on thermal hydraulic effect occurring in the containment. It is thought that condensation of steam included in the hydrogen mixture is very important in the hydrogen behaviors.

  14. 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)

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

    Directory of Open Access Journals (Sweden)

    M.H. Altaf

    2014-12-01

    Full Text Available Burnup dependent steady state thermal hydraulic analysis of TRIGA Mark-II research reactor has been carried out utilizing coupled point kinetics, neutronics and thermal hydraulics code EUREKA-2/RR. From the previous calculations of neutronics parameters including percentage burnup of individual fuel elements performed so far for 700 MWD burnt core of TRIGA reactor showed that the fuel rod predicted as hottest at the beginning of cycle (fresh core was found to remain as the hottest until 200 MWD of burn, but, with the progress of core burn, the hottest rod was found to be shifted and another rod in the core became the hottest. The present study intends to evaluate the thermal hydraulic parameters of these hottest fuel rods at different cycles of burnup, from beginning to 700 MWD core burnt considering reactor operates under steady state condition. Peak fuel centerline temperature, maximum cladding and coolant temperatures of the hottest channels were calculated. It revealed that maximum temperature reported for fuel clad and fuel centerline found to lie below their melting points which indicate that there is no chance of burnout on the fuel cladding surface and no blister in the fuel meat throughout the considered cycles of core burnt.

  16. Effect of injection screen slot geometry on hydraulic conductivity tests

    Science.gov (United States)

    Klammler, Harald; Nemer, Bassel; Hatfield, Kirk

    2014-04-01

    Hydraulic conductivity and its spatial variability are important hydrogeological parameters and are typically determined through injection tests at different scales. For injection test interpretation, shape factors are required to account for injection screen geometry. Shape factors act as proportionality constants between hydraulic conductivity and observed ratios of injection flow rate and injection head at steady-state. Existing results for such shape factors assume either an ideal screen (i.e., ignoring effects of screen slot geometry) or infinite screen length (i.e., ignoring effects of screen extremes). In the present work, we investigate the combined effects of circumferential screen slot geometry and finite screen length on injection shape factors. This is done in terms of a screen entrance resistance by solving a steady-state potential flow mixed type boundary value problem in a homogeneous axi-symmetric flow domain using a semi-analytical solution approach. Results are compared to existing analytical solutions for circumferential and longitudinal slots on infinite screens, which are found to be identical. Based on an existing approximation, an expression is developed for a dimensionless screen entrance resistance of infinite screens, which is a function of the relative slot area only. For anisotropic conditions, e.g., when conductivity is smaller in the vertical direction than in the horizontal, screen entrance losses for circumferential slots increase, while they remain unaffected for longitudinal slots. This work is not concerned with investigating the effects of (possibly turbulent) head losses inside the injection device including the passage through the injection slots prior to entering the porous aquifer.

  17. Multipurpose Thermal Insulation Test Apparatus

    Science.gov (United States)

    Fesmire, James E. (Inventor); Augustynowicz, Stanislaw D. (Inventor)

    2002-01-01

    A multi-purpose thermal insulation test apparatus is used for testing insulation materials, or other components. The test apparatus is a fluid boil-off calorimeter system for calibrated measurement of the apparent thermal conductivity (k-value) of a specimen material at a fixed vacuum level. The apparatus includes an inner vessel for receiving a fluid with a normal boiling point below ambient temperature, such as liquid nitrogen, enclosed within a vacuum chamber. A cold mass assembly, including the inner vessel and thermal guards, is suspended from the top of the vacuum chamber. Handling tools attach to the cold mass assembly for convenient manipulation of the assembly and for the installation or wrapping of insulation test materials. Liquid nitrogen is typically supplied to the inner vessel using a fill tube with funnel. A single port through the top of the vacuum chamber facilitates both filling and venting. Aerogel composite stacks with reflective films are fastened to the top and the bottom of the inner vessel as thermal guards. The comparative k-value of the insulation material is determined by measuring the boil-off flow rate of gas, the temperature differential across the insulation thickness, and the dimensions (length and diameters) of the test specimen.

  18. Hydraulic tests for the Excavation Disturbed Zone in NATM drift of North Extension

    Energy Technology Data Exchange (ETDEWEB)

    Matsuoka, Eiken [Taisei Kiso Sekkei Co., Ltd., Tokyo (Japan)

    1997-03-01

    Investigation for characterization of rock properties of the Excavation Disturbed Zone (EDZ) were carried out in NATM drift of North Extension in the Tono Mine. As a part of this investigation, hydraulic tests were performed by means of the hydraulic measuring instrument, which had been developed by PNC Tono Geoscience Center. The purpose of this tests is to characterize the change in hydraulic properties of the EDZ caused by drift excavation using machine (boom header). The hydraulic tests were performed in the burials MH-1,2,3, in which hydraulic tests had been performed before the drift excavation in 1994. The test results indicate that the measured values of pore water pressure have decreased after excavation of the drift. The values ranged from -0.037 kgf/cm{sup 2} to 0.039 kgf/cm{sup 2}. The measured hydraulic conductivities ranged from 2.2*10{sup -11} cm/s to 9.1*10{sup -11} cm/s for mud stone and from 2.8*10{sup -9} cm/s to 2.4*10{sup -7} cm/s for conglomerate. The measured hydraulic conductivities for mud stone are below the lower limit of the instrument, and the change in the hydraulic conductivities for conglomerate is little. The hydraulic conductivities for conglomerate and mad stone (reference values) are interpreted. The change in hydraulic conductivities measured before and after excavation of the drift is insignificant. (author)

  19. Stomatal control and leaf thermal and hydraulic capacitances under rapid environmental fluctuations.

    Directory of Open Access Journals (Sweden)

    Stanislaus J Schymanski

    Full Text Available Leaves within a canopy may experience rapid and extreme fluctuations in ambient conditions. A shaded leaf, for example, may become exposed to an order of magnitude increase in solar radiation within a few seconds, due to sunflecks or canopy motions. Considering typical time scales for stomatal adjustments, (2 to 60 minutes, the gap between these two time scales raised the question whether leaves rely on their hydraulic and thermal capacitances for passive protection from hydraulic failure or over-heating until stomata have adjusted. We employed a physically based model to systematically study effects of short-term fluctuations in irradiance on leaf temperatures and transpiration rates. Considering typical amplitudes and time scales of such fluctuations, the importance of leaf heat and water capacities for avoiding damaging leaf temperatures and hydraulic failure were investigated. The results suggest that common leaf heat capacities are not sufficient to protect a non-transpiring leaf from over-heating during sunflecks of several minutes duration whereas transpirative cooling provides effective protection. A comparison of the simulated time scales for heat damage in the absence of evaporative cooling with observed stomatal response times suggested that stomata must be already open before arrival of a sunfleck to avoid over-heating to critical leaf temperatures. This is consistent with measured stomatal conductances in shaded leaves and has implications for water use efficiency of deep canopy leaves and vulnerability to heat damage during drought. Our results also suggest that typical leaf water contents could sustain several minutes of evaporative cooling during a sunfleck without increasing the xylem water supply and thus risking embolism. We thus submit that shaded leaves rely on hydraulic capacitance and evaporative cooling to avoid over-heating and hydraulic failure during exposure to typical sunflecks, whereas thermal capacitance provides

  20. Study of the thermal and hydraulic phenomena occurring during power excursion on a heated test section; Etude des phenomenes thermiques et hydrauliques accompagnant une excursion rapide de puissance sur un canal chauffant

    Energy Technology Data Exchange (ETDEWEB)

    Nyer, M. [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1969-07-01

    The thermal and hydrodynamic phenomena occurring during a power excursion were studied in an out-of-pile loop with a water cooled channel at low pressure (1 to 4 atm. abs. ). Circular and rectangular test sections with electrically heated walls of two different thermal diffusivity materials(aluminium and stainless steel) were used. The rectangular test sections were 600 mm long, 35 mm wide and had a 2, 9 mm gap; they simulate two half plates of the M.T.R. fuel element. Natural or forced convection are possible in the test section; the water height above it can be varied from 2.8 to 8 meters and the maximum allowed pressure at its outlet is 4 atm. abs.The heating source is a series of lead batteries which is able to generate, for short periods of time, 85 volts and 25000 amperes; linear, square or exponential power rise versus time can be realized. A 14 channels tape recorder (0-10 000 Hz bandwidth; is used for the measurements of temperature (8/100 mm diameter thermocouple), pressure ('Statham' pressure transducers) and void fraction (X rays). More than 500 tests have been carried out. The influence of the initial water temperature, flow rate, pressure, water height on the water ejections, pressure variations and void fraction in the test section were studied. Tests with energies up to 3000 W/cm in 50 milliseconds were attempted. The energy above which the instabilities appear was determined. An interpretation of the observed phenomena and a simplified theoretical model are presented. [French] Les phenomenes thermiques et hydrodynamiques qui apparaissent au cours d'une excursion de puissance ont ete etudies sur un canal refroidi par de l'eau a basse pression situe sur une installation hors pile. On a utilise des sections d'essais de geometrie cylindrique ou parallipedique dont les parois chauffees par effet Joule sont constituees de materiaux de diffusivite calorifique differente (aluminium et acier inoxydable). La section d

  1. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  2. Use of a hydraulic brake as a source of thermal energy for the railway rolling stock

    Directory of Open Access Journals (Sweden)

    V.A.Gabrinets

    2012-12-01

    Full Text Available Introduction: In this paper the braking issues of passenger trains which have a great speed and frequent stops are examines. Problem statement: These processes are ехpensive and have big energy losses. The proposed solution to the problem: The kinetic energy of braking prosses propose to turn into thermal energy of heating fluid. For this purpose special hydraulic brake is proposed. The brake is connected with the wheel carriage pairs. The process is based on the energy dissipation in liqid when the disks with spikes rotate in it. Because the real liquid has friction and viscosity, it will be heat up, when the mechanical parts of the hydraulic brake are moved in it. The design, operating principle and characteristics of the hydraulic brake are proposed. Transmission of kinetic energy of carriage motion to brake system executed by mechanical clutches. It connected with the wheel pair and transmitting the energy the wheels rotation to hydraulic brake discs. The cylindrical rods are installed on the discs. Rods location fits the profile of the curved centrifugal pump vanes. As result, the fluid heatind prosess by rotatinge discs with rods take place also at the same time with the liquid pumping through the inner volume of brake system.Conclusions: Affordable passenger carriage braking dynamic is achieved by varying the size and number of rods. The heated liquid may be subsequently used for household needs and for heating the passenger carriage.

  3. TAPINS: A THERMAL-HYDRAULIC SYSTEM CODE FOR TRANSIENT ANALYSIS OF A FULLY-PASSIVE INTEGRAL PWR

    Directory of Open Access Journals (Sweden)

    YEON-GUN LEE

    2013-08-01

    Full Text Available REX-10 is a fully-passive small modular reactor in which the coolant flow is driven by natural circulation, the RCS is pressurized by a steam-gas pressurizer, and the decay heat is removed by the PRHRS. To confirm design decisions and analyze the transient responses of an integral PWR such as REX-10, a thermal-hydraulic system code named TAPINS (Thermal-hydraulic Analysis Program for INtegral reactor System is developed in this study. Based on a one-dimensional four-equation drift-flux model, TAPINS incorporates mathematical models for the core, the helical-coil steam generator, and the steam-gas pressurizer. The system of difference equations derived from the semi-implicit finite-difference scheme is numerically solved by the Newton Block Gauss Seidel (NBGS method. TAPINS is characterized by applicability to transients with non-equilibrium effects, better prediction of the transient behavior of a pressurizer containing non-condensable gas, and code assessment by using the experimental data from the autonomous integral effect tests in the RTF (REX-10 Test Facility. Details on the hydrodynamic models as well as a part of validation results that reveal the features of TAPINS are presented in this paper.

  4. Open stack thermal battery tests

    Energy Technology Data Exchange (ETDEWEB)

    Long, Kevin N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Roberts, Christine C. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Grillet, Anne M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Headley, Alexander J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Fenton, Kyle [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wong, Dennis [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ingersoll, David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-04-17

    We present selected results from a series of Open Stack thermal battery tests performed in FY14 and FY15 and discuss our findings. These tests were meant to provide validation data for the comprehensive thermal battery simulation tools currently under development in Sierra/Aria under known conditions compared with as-manufactured batteries. We are able to satisfy this original objective in the present study for some test conditions. Measurements from each test include: nominal stack pressure (axial stress) vs. time in the cold state and during battery ignition, battery voltage vs. time against a prescribed current draw with periodic pulses, and images transverse to the battery axis from which cell displacements are computed. Six battery configurations were evaluated: 3, 5, and 10 cell stacks sandwiched between 4 layers of the materials used for axial thermal insulation, either Fiberfrax Board or MinK. In addition to the results from 3, 5, and 10 cell stacks with either in-line Fiberfrax Board or MinK insulation, a series of cell-free “control” tests were performed that show the inherent settling and stress relaxation based on the interaction between the insulation and heat pellets alone.

  5. Sensitivity analysis of hydraulic and thermal parameters inducing anomalous heat flow in the Lower Yarmouk Gorge

    Science.gov (United States)

    Goretzki, Nora; Inbar, Nimrod; Kühn, Michael; Möller, Peter; Rosenthal, Eliyahu; Schneider, Michael; Siebert, Christian; Magri, Fabien

    2016-04-01

    The Lower Yarmouk Gorge, at the border between Israel and Jordan, is characterized by an anomalous temperature gradient of 46 °C/km. Numerical simulations of thermally-driven flow show that ascending thermal waters are the result of mixed convection, i.e. the interaction between the regional flow from the surrounding heights and buoyant flow within permeable faults [1]. Those models were calibrated against available temperature logs by running several forward problems (FP), with a classic "trial and error" method. In the present study, inverse problems (IP) are applied to find alternative parameter distributions that also lead to the observed thermal anomalies. The investigated physical parameters are hydraulic conductivity and thermal conductivity. To solve the IP, the PEST® code [2] is applied via the graphical interface FEPEST® in FEFLOW® [3]. The results show that both hydraulic and thermal conductivity are consistent with the values determined with the trial and error calibrations, which precede this study. However, the IP indicates that the hydraulic conductivity of the Senonian Paleocene aquitard can be 8.54*10-3 m/d, which is three times lower than the originally estimated value in [1]. Moreover, the IP suggests that the hydraulic conductivity in the faults can increase locally up to 0.17 m/d. These highly permeable areas can be interpreted as local damage zones at the faults/units intersections. They can act as lateral pathways in the deep aquifers that allow deep outflow of thermal water. This presentation provides an example about the application of FP and IP to infer a wide range of parameter values that reproduce observed environmental issues. [1] Magri F, Inbar N, Siebert C, Rosenthal E, Guttman J, Möller P (2015) Transient simulations of large-scale hydrogeological processes causing temperature and salinity anomalies in the Tiberias Basin. Journal of Hydrology, 520, 342-355 [2] Doherty J (2010) PEST: Model-Independent Parameter Estimation. user

  6. 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)

  7. In-Plant Testing of High-Efficiency Hydraulic Separators

    Energy Technology Data Exchange (ETDEWEB)

    G. H. Luttrell; R. Q. Honaker; R. C. Bratton; T. C. Westerfield; J. N. Kohmuench

    2004-07-20

    The mineral processing industry has commonly utilized hydraulic separators throughout history for classification and gravity concentration of various minerals. More commonly referred to as hindered-bed or fluidized-bed separators, these units make use of differential particle settling rates to segregate particles according to shape, size, and/or density. As with any equipment, there are inefficiencies associated with its operation, which prompted an industry driven research program to further evaluate two novel high-efficiency hindered bed separators. These units, which are commercially called the CrossFlow separator and HydroFloat separator, have the potential to improve performance (separation efficiency and throughput) and reduce operating costs (power consumption, water and reagent usage). This report describes the results of Phase I activities (laboratory and pilot-scale tests) conducted with the CrossFlow and HydroFloat separators at several locations in the minerals and coal industries. Details of the testing programs (equipment setup, shakedown testing and detailed testing) associated with four coal plants and two phosphate plants are summarized in this work. In most of these applications, the high-efficiency units proved to provide a higher quality product at reduced costs when compared against the performance of conventional separators. Based on promising results obtained from Phase I, full-scale prototypes will be purchased by several mining companies for use in Phase II of this project. Two of the prototype units, which will be constructed by Eriez Manufacturing, are expected to be installed by a major U.S. phosphate producer and a large eastern U.S. coal company. Negotiations are also underway to purchase and install additional prototype units by a mineral sands producer and a second phosphate producer. The data obtained from the full-scale evaluations will be used to further promote commercialization and industrial applications of these innovative

  8. Transient analysis in the 3D nodal kinetics and thermal-hydraulics ANDES/COBRA coupled system

    Energy Technology Data Exchange (ETDEWEB)

    Lozano, Juan Andres; Aragones, Jose Maria; Garcia-Herranz, Nuria [Universidad Politecnica de Madrid, Madrid (Spain)

    2008-07-01

    Neutron kinetics has been implemented in the 3D nodal solver ANDES, which has been coupled to the core thermal-hydraulics (TH) code COBRA-III for core transient analysis. The purpose of this work is, first, to discuss and test the ability of the kinetics solver ANDES to model transients; and second, by means of a systematic analysis, including alternate kinetics schemes, time step size, nodal size, neutron energy groups and spectrum, to serve as a basis for the development of more accurate and efficient neutronics/thermal-hydraulics tools for general transient simulations. The PWR MOX/UO{sub 2} transient benchmark provided by the OECD/NEA and US NRC was selected for these goals. The obtained ANDES/COBRA-III results were consistent with other solutions to the benchmark; the differences in the TH feedback led to slight differences in the core power evolution, whereas very good agreements were found in the other requested parameters. The performed systematic analysis highlighted the optimum kinetics iterative scheme, and showed that neutronics spatial discretization effects have stronger influence than time discretization effects, in the semi-implicit scheme adopted, on the numerical solution. On the other hand, the number of energy groups has an important influence on the transient evolution, whereas the assumption of using the prompt neutron spectrum for delayed neutrons is acceptable as it leads to small relative errors. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Podlazov, L. N.

    1998-07-29

    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.

  10. 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

  11. A General Model for Thermal, Hydraulic and Electric Analysis of Superconducting Cables

    CERN Document Server

    Bottura, L; Rosso, C

    2000-01-01

    In this paper we describe a generic, multi-component and multi-channel model for the analysis of superconducting cables. The aim of the model is to treat in a general and consistent manner simultaneous thermal, electric and hydraulic transients in cables. The model is devised for most general situations, but reduces in limiting cases to most common approximations without loss of efficiency. We discuss here the governing equations, and we write them in a matrix form that is well adapted to numerical treatment. We finally demonstrate the model capability by comparison with published experimental data on current distribution in a two-strand cable.

  12. Neutronic and thermal-hydraulic analysis of fission molybdenum-99 production at Tehran Research Reactor using LEU plate targets.

    Science.gov (United States)

    Abedi, Ebrahim; Ebrahimkhani, Marzieh; Davari, Amin; Mirvakili, Seyed Mohammad; Tabasi, Mohsen; Maragheh, Mohammad Ghannadi

    2016-12-01

    Efficient and safe production of molybdenum-99 ( 99 Mo) radiopharmaceutical at Tehran Research Reactor (TRR) via fission of LEU targets is studied. Neutronic calculations are performed to evaluate produced 99 Mo activity, core neutronic safety parameters and also the power deposition values in target plates during a 7 days irradiation interval. Thermal-hydraulic analysis has been also carried out to obtain thermal behavior of these plates. Using Thermal-hydraulic analysis, it can be concluded that the safety parameters are satisfied in the current study. Consequently, the present neutronic and thermal-hydraulic calculations show efficient 99 Mo production is accessible at significant activity values in TRR current core configuration. Copyright © 2016 Elsevier Ltd. All rights reserved.

  13. Thermal-hydraulics/thermal-mechanics temporal coupling for unprotected loss of flow accidents simulations on a SFR

    Directory of Open Access Journals (Sweden)

    Patricot Cyril

    2016-01-01

    Full Text Available In the frame of ASTRID designing, unprotected loss of flow (ULOF accidents are considered. As the reactor is not scrammed, power evolution is driven by neutronic feedbacks, among which Doppler effect, linked to fuel temperature, is prominent. Fuel temperature is calculated using thermal properties of fuel pins (we will focus on heat transfer coefficient between fuel pellet and cladding, Hgap, and on fuel thermal conductivity, λfuel which vary with irradiation conditions (neutronic flux, mass flow and history for instance and during transient (mainly because of dilatation of materials with temperature. In this paper, we propose an analysis of the impact of spatial variation and temporal evolution of thermal properties of fuel pins on a CFV-like core [M.S. Chenaud et al., Status of the ASTRID core at the end of the pre-conceptual design phase 1, in Proceedings of ICAPP 2013, Jeju Island, Korea (2013] behavior during an ULOF accident. These effects are usually neglected under some a priori conservative assumptions. The vocation of our work is not to provide a best-estimate calculation of ULOF transient, but to discuss some of its physical aspects. To achieve this goal, we used TETAR, a thermal-hydraulics system code developed by our team to calculate ULOF transients, GERMINAL V1.5, a CEA code dedicated to SFR pin thermal-mechanics calculations and APOLLO3®, a neutronic code in development at CEA.

  14. Thermal Hydraulics Design and Analysis Methodology for a Solid-Core Nuclear Thermal Rocket Engine Thrust Chamber

    Science.gov (United States)

    Wang, Ten-See; Canabal, Francisco; Chen, Yen-Sen; Cheng, Gary; Ito, Yasushi

    2013-01-01

    Nuclear thermal propulsion is a leading candidate for in-space propulsion for human Mars missions. This chapter describes a thermal hydraulics design and analysis methodology developed at the NASA Marshall Space Flight Center, in support of the nuclear thermal propulsion development effort. The objective of this campaign is to bridge the design methods in the Rover/NERVA era, with a modern computational fluid dynamics and heat transfer methodology, to predict thermal, fluid, and hydrogen environments of a hypothetical solid-core, nuclear thermal engine the Small Engine, designed in the 1960s. The computational methodology is based on an unstructured-grid, pressure-based, all speeds, chemically reacting, computational fluid dynamics and heat transfer platform, while formulations of flow and heat transfer through porous and solid media were implemented to describe those of hydrogen flow channels inside the solid24 core. Design analyses of a single flow element and the entire solid-core thrust chamber of the Small Engine were performed and the results are presented herein

  15. 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)

  16. Thermal-hydraulic analysis techniques for axisymmetric pebble bed nuclear reactor cores. [PEBBLE code

    Energy Technology Data Exchange (ETDEWEB)

    Stroh, K.R.

    1979-03-01

    The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases.

  17. MyrrhaFoam: A CFD model for the study of the thermal hydraulic behavior of MYRRHA

    Energy Technology Data Exchange (ETDEWEB)

    Koloszar, Lilla; Buckingham, Sophia; Planquart, Philippe [von Karman Institute, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Keijers, Steven [SCK-CEN, Boeretang 200, 2400 Mol (Belgium)

    2017-02-15

    Highlights: • Development of a modeling approach for simulating the thermal hydraulics of heavy liquid metal nuclear reactors. • Detailed description of the modeling of each component through the MYRRHA reactor. • Detailed analysis of the flow field of the MYRRHA reactor under operating condition. • Assessment of the thermal load on the structures as well as the thermal stratification in the upper and the lower plenum. - Abstract: Numerical analysis of the thermohydraulic behavior of the innovative flexible fast spectrum research reactor, MYRRHA, under design by the Belgian Nuclear Research Center (SCK• CEN) is a very challenging task. The primary coolant of the reactor is Lead Bismuth Eutectic, LBE, which is an opaque heavy liquid metal with low Prandtl number. The simulation tool needs to involve many complex physical phenomena to be able to predict accurately the flow and thermal field in the pool type reactor. In the past few years, within the frame of a collaboration between SCK• CEN and the von Karman Institute, a new platform, MyrrhaFoam, was developed based on the open source simulation environment, OpenFOAM. The current tool can deal with incompressible buoyancy corrected steady/unsteady single phase flows. It takes into account conjugate heat transfer in the solid parts which is mandatory due to the expected high temperature gradients between the different parts of the reactor. The temperature dependent properties of LBE are also considered. MyrrhaFoam is supplemented with the most relevant thermal turbulence models for low Prandtl number liquids up to date.

  18. 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.

  19. 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.

  20. 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)

  1. IMPROVEMENT OF THE LOCA PSA MODEL USING A BEST-ESTIMATE THERMAL-HYDRAULIC ANALYSIS

    Directory of Open Access Journals (Sweden)

    DONG HYUN LEE

    2014-08-01

    Full Text Available Probabilistic Safety Assessment (PSA has been widely used to estimate the overall safety of nuclear power plants (NPP and it provides base information for risk informed application (RIA and risk informed regulation (RIR. For the effective and correct use of PSA in RIA/RIR related decision making, the risk estimated by a PSA model should be as realistic as possible. In this work, a best-estimate thermal-hydraulic analysis of loss-of-coolant accidents (LOCAs for the Hanul Nuclear Units 3&4 is first carried out in a systematic way. That is, the behaviors of peak cladding temperature (PCT were analyzed with various combinations of break sizes, the operating conditions of safety systems, and the operator's action time for aggressive secondary cooling. Thereafter, the results of the thermal-hydraulic analysis have been reflected in the improvement of the PSA model by changing both accident sequences and success criteria of the event trees for the LOCA scenarios.

  2. 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)

  3. 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)

  4. Porosity Effect in the Core Thermal Hydraulics for Ultra High Temperature Gas-cooled Reactor

    Directory of Open Access Journals (Sweden)

    Motoo Fumizawa

    2008-12-01

    Full Text Available This study presents an experimental method of porosity evaluation and a predictive thermal-hydraulic analysis with packed spheres in a nuclear reactor core. The porosity experiments were carried out in both a fully shaken state with the closest possible packing and in a state of non-vibration. The predictive analysis considering the fixed porosity value was applied as a design condition for an Ultra High Temperature Reactor Experiment (UHTREX. The thermal-hydraulic computer code was developed and identified as PEBTEMP. The highest outlet coolant temperature of 1316 oC was achieved in the case of an UHTREX at Los Alamos Scientific Laboratory, which was a small scale UHTR. In the present study, the fuel was changed to a pebble type, a porous media. In order to compare the present pebble bed reactor and UHTREX, a calculation based on HTGR-GT300 was carried out in similar conditions with UHTREX; in other words, with an inlet coolant temperature of 871oC, system pressure of 3.45 MPa and power density of 1.3 w/cm3. As a result, the fuel temperature in the present pebble bed reactor showed an extremely lower value compared to that of UHTREX.

  5. Comparative study of boron transport models in NRC Thermal-Hydraulic Code Trace

    Energy Technology Data Exchange (ETDEWEB)

    Olmo-Juan, Nicolás; Barrachina, Teresa; Miró, Rafael; Verdú, Gumersindo; Pereira, Claubia, E-mail: nioljua@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es, E-mail: claubia@nuclear.ufmg.br [Institute for Industrial, Radiophysical and Environmental Safety (ISIRYM). Universitat Politècnica de València (Spain); Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    Recently, the interest in the study of various types of transients involving changes in the boron concentration inside the reactor, has led to an increase in the interest of developing and studying new models and tools that allow a correct study of boron transport. Therefore, a significant variety of different boron transport models and spatial difference schemes are available in the thermal-hydraulic codes, as TRACE. According to this interest, in this work it will be compared the results obtained using the different boron transport models implemented in the NRC thermal-hydraulic code TRACE. To do this, a set of models have been created using the different options and configurations that could have influence in boron transport. These models allow to reproduce a simple event of filling or emptying the boron concentration in a long pipe. Moreover, with the aim to compare the differences obtained when one-dimensional or three-dimensional components are chosen, it has modeled many different cases using only pipe components or a mix of pipe and vessel components. In addition, the influence of the void fraction in the boron transport has been studied and compared under close conditions to BWR commercial model. A final collection of the different cases and boron transport models are compared between them and those corresponding to the analytical solution provided by the Burgers equation. From this comparison, important conclusions are drawn that will be the basis of modeling the boron transport in TRACE adequately. (author)

  6. 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.

  7. Thermal-hydraulic criteria for the APT tungsten neutron source design

    Energy Technology Data Exchange (ETDEWEB)

    Pasamehmetoglu, K.

    1998-03-01

    This report presents the thermal-hydraulic design criteria (THDC) developed for the tungsten neutron source (TNS). The THDC are developed for the normal operations, operational transients, and design-basis accidents. The requirements of the safety analyses are incorporated into the design criteria, consistent with the integrated safety management and the safety-by-design philosophy implemented throughout the APT design process. The phenomenology limiting the thermal-hydraulic design and the confidence level requirements for each limit are discussed. The overall philosophy of the uncertainty analyses and the confidence level requirements also are presented. Different sets of criteria are developed for normal operations, operational transients, anticipated accidents, unlikely accidents, extremely unlikely accidents, and accidents during TNS replacement. In general, the philosophy is to use the strictest criteria for the high-frequency events. The criteria is relaxed as the event frequencies become smaller. The THDC must be considered as a guide for the design philosophy and not as a hard limit. When achievable, design margins greater than those required by the THDC must be used. However, if a specific event sequence cannot meet the THDC, expensive design changes are not necessary if the single event sequence results in sufficient margin to safety criteria and does not challenge the plant availability or investment protection considerations.

  8. Development of an integrated thermal-hydraulics capability incorporating RELAP5 and PANTHER neutronics code

    Energy Technology Data Exchange (ETDEWEB)

    Page, R.; Jones, J.R.

    1997-07-01

    Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell `B` Loss of offsite power fault transient.

  9. Multi channel thermal hydraulic analysis of gas cooled fast reactor using genetic algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Drajat, R. Z.; Su' ud, Z.; Soewono, E.; Gunawan, A. Y. [Department of Mathematics, Institut Teknologi Bandung, Bandung 40132 (Indonesia); Department of Physics, Institut Teknologi Bandung, Bandung 40132 (Indonesia); Department of Mathematics, Institut Teknologi Bandung, Bandung 40132 (Indonesia)

    2012-05-22

    There are three analyzes to be done in the design process of nuclear reactor i.e. neutronic analysis, thermal hydraulic analysis and thermodynamic analysis. The focus in this article is the thermal hydraulic analysis, which has a very important role in terms of system efficiency and the selection of the optimal design. This analysis is performed in a type of Gas Cooled Fast Reactor (GFR) using cooling Helium (He). The heat from nuclear fission reactions in nuclear reactors will be distributed through the process of conduction in fuel elements. Furthermore, the heat is delivered through a process of heat convection in the fluid flow in cooling channel. Temperature changes that occur in the coolant channels cause a decrease in pressure at the top of the reactor core. The governing equations in each channel consist of mass balance, momentum balance, energy balance, mass conservation and ideal gas equation. The problem is reduced to finding flow rates in each channel such that the pressure drops at the top of the reactor core are all equal. The problem is solved numerically with the genetic algorithm method. Flow rates and temperature distribution in each channel are obtained here.

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

    Science.gov (United States)

    Domalapally, Phani; Di Caro, Marco

    2017-11-01

    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. Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Cui, Shijie; Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn; Cheng, Jie; Tian, Wenxi; Su, G.H.

    2017-01-15

    As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

  12. The influence of the equivalent hydraulic diameter on the pressure drop prediction of annular test section

    Science.gov (United States)

    Al-Kayiem, A. H. H.; Ibrahim, M. A.

    2015-12-01

    The flow behaviour and the pressure drop throughout an annular flow test section was investigated in order to evaluate and justify the reliability of experimental flow loop for wax deposition studies. The specific objective of the present paper is to assess and highlight the influence of the equivalent diameter method on the analysis of the hydrodynamic behaviour of the flow and the pressure drop throughout the annular test section. The test section has annular shape of 3 m length with three flow passages, namely; outer thermal control jacket, oil annular flow and inner pipe flow of a coolant. The oil annular flow has internal and external diameters of 0.0422 m and 0.0801 m, respectively. Oil was re-circulated in the annular passage while a cold water-glycol mixture was re-circulated in the inner pipe counter currently to the oil flow. The experiments were carried out at oil Reynolds number range of 2000 to 17000, covering laminar, transition and turbulent flow regimes. Four different methods of equivalent diameter of the annulus have been considered in this hydraulic analysis. The correction factor model for frictional pressure drop was also considered in the investigations. All methods addressed the high deviation of the prediction from the experimental data, which justified the need of a suitable pressure prediction correlation for the annular test section. The conventional hydraulic diameter method is a convenient substitute for characterizing physical dimension of a non-circular duct, and it leads to fairly good correlation between turbulent fluid flow and heat transfer characteristic of annular ducts.

  13. Model of dynamic compression tests on hydraulic testing machines: Influence of dynamic phenomena

    Science.gov (United States)

    Diot, S.; Gavrus, A.; Guines, D.; Ragneau, E.

    2003-09-01

    The forming process simulation requires models describing the materials behaviour at large strains and at strain rates up to hundreds of s^{-1}. The major difficulty then encountered is that few experimental tests enable to reach these two criteria. For a few years, several studies have been carried out on hydraulic machines provided with a dynamic jack. However, for higher strain rates tests, the load measurement is disturbed by the response of the experimental set-up and oscillations appear. In this article, the experimental test is developed and a finite element model of the set-up is introduced.

  14. Reservoir Stimulation Experiments at the Grimsel Test Site: Stress Measurements using Hydraulic fracturing, Hydraulic Tests on Pre-existing Fractures and Overcoring

    Science.gov (United States)

    Doetsch, J.; Gischig, V.; Amann, F.; Madonna, C.; Jalali, M.; Valley, B.; Evans, K. F.

    2015-12-01

    A decameter-scale in-situ hydraulic stimulation and circulation experiment has been planned in the Deep Underground rock Laboratory (DUG Lab) at the Grimsel Test Site, Switzerland. The general objective of this experiment is to improve our understanding of the pressure, temperature and stress changes in the rock mass due to hydraulic stimulation. In this context, the main goal is to investigate the effect of hydro-shearing on the local stress variation as well as transient and permanent permeability changes with comprehensive thermo-hydro-mechanical (THM) and acoustic emission monitoring. This experiment is designed such that stimulation processes are recorded in a dataset that is unique in THM coupled processes and induced seismicity research. In preparation to the hydro-shearing experiments, the experimental rock volume has been studied in detail using geological tunnel mapping, optical televiewer in existing boreholes, hydraulic tests, geophysical imaging and review of the extensive literature on experiments at the Grimsel Test Site. The geophysical investigations include reflection and transmission ground penetrating radar (GPR) and seismic measurements between the tunnels to image shear zones and reveal heterogeneity of the rock mass. The orientation and magnitude of the principal stresses of the rock volume and its surroundings has been analyzed using hydraulic fracturing, hydraulic tests on pre-existing fractures and overcoring. The hydraulic fracturing tests for stress measurements were monitored using a 32-channel acoustic emission monitoring system and a regional seismic monitoring network. Here, we present the results of the pre-investigations and stress measurements, and give an outlook for the hydro-shearing experiments planned for spring 2016.

  15. Development of the thermal hydraulic analysis code for a copper bonded steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Kim, W. K.; Wei, M. H.; Yeo, J. H.; Kim, S. O. [KAERI, Taejon (Korea, Republic of); Back, B. J. [Chonbuk National Univ., Jeonju (Korea, Republic of)

    2002-10-01

    An one-dimensional thermal-hydraulic analysis computer code was developed for the thermal sizing of copper bonded steam generator. It was assumed that the conduction heat transfer of copper region between hot side and cold side tube is one-dimensional and its thermal resistance of the function of a tube pitch was derived. The flow regions of water/steam side were devided into four regions, which are sub-cooled, saturated, film boiling, and super-heated regions. The numbers of tube were selected from 250 to 3500 for the parameter study calculation. The pitch over tube diameter ratios were 1.4, 1.6 and 1.8. The calculation results showed that when the number of tube was 2500, the length of heating tube was about 10 m and the diameter was about 3 m. If P/D ratio increases, the thermal resistance of copper component also increases, however the length of heating tube is not increasing so much.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  17. An analytical study on excitation of nuclear-coupled thermal-hydraulic instability due to seismically induced resonance in BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Masashi [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    1997-07-01

    This paper describes the results of a scoping study on seismically induced resonance of nuclear-coupled thermal-hydraulic instability in BWRs, which was conducted by using TRAC-BF1 within a framework of a point kinetics model. As a result of the analysis, it is shown that a reactivity insertion could occur accompanied by in-surge of coolant into the core resulted from the excitation of the nuclear-coupled instability by the external acceleration. In order to analyze this phenomenon more in detail, it is necessary to couple a thermal-hydraulic code with a three-dimensional nuclear kinetics code.

  18. How do trees die? A test of the hydraulic failure and carbon starvation hypotheses.

    Science.gov (United States)

    Sevanto, Sanna; McDowell, Nate G; Dickman, L Turin; Pangle, Robert; Pockman, William T

    2014-01-01

    Despite decades of research on plant drought tolerance, the physiological mechanisms by which trees succumb to drought are still under debate. We report results from an experiment designed to separate and test the current leading hypotheses of tree mortality. We show that piñon pine (Pinus edulis) trees can die of both hydraulic failure and carbon starvation, and that during drought, the loss of conductivity and carbohydrate reserves can also co-occur. Hydraulic constraints on plant carbohydrate use determined survival time: turgor loss in the phloem limited access to carbohydrate reserves, but hydraulic control of respiration prolonged survival. Our data also demonstrate that hydraulic failure may be associated with loss of adequate tissue carbohydrate content required for osmoregulation, which then promotes failure to maintain hydraulic integrity. Published 2013. This article is a U.S. Government work and is in the public domain in the USA.

  19. Spatial variability of hydraulic conductivity of an unconfined sandy aquifer determined by a mini slug test

    DEFF Research Database (Denmark)

    Bjerg, Poul Løgstrup; Hinsby, Klaus; Christensen, Thomas Højlund

    1992-01-01

    The spatial variability of the hydraulic conductivity in a sandy aquifer has been determined by a mini slug test method. The hydraulic conductivity (K) of the aquifer has a geometric mean of 5.05 × 10−4 m s−1, and an overall variance of 1n K equal to 0.37 which corresponds quite well to the results...... obtained by two large scale tracer experiments performed in the aquifer. A geological model of the aquifer based on 31 sediment cores, proposed three hydrogeological layers in the aquifer concurrent with the vertical variations observed with respect to hydraulic conductivity. The horizontal correlation...... length of the hydraulic conductivity has been determined for each of the three hydrogeological layers and is found to be small (1–2.5 m). The asymptotic longitudinal dispersivity of the aquifer has been estimated from the variance in hydraulic conductivity and the horizontal correlation length...

  20. Model of a nuclear thermal test pipe using ATHENA

    Energy Technology Data Exchange (ETDEWEB)

    Dibben, Mark J. [Air Force Inst. of Technology, Wright-Patterson AFB (United States)

    1992-03-01

    Nuclear thermal propulsion offers significant improvements in rocket engine specific impulse over rockets employing chemical propulsion. The computer code ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) was used in a parametric analysis of a fuelpipe. The fuelpipe is an annular particle bed fuel element of the reactor with radially inward flow of hydrogen through it. The outlet temperature of the hydrogen is parametrically related to key effects, including the effect of reactor power at two different pressure drops, the effect of the power coupling factor of the Annular Core Research Reactor, and the effect of hydrogen flow. Results show that the outlet temperature is linearly related to the reactor power and nonlinearly to the change in pressure drop. The linear relationship at higher temperatures is probably not valid due to dissociation of hydrogen. Once thermal properties of hydrogen become available, the ATHENA model for this study could easily be modified to test this conjecture.

  1. Thermal Properties for the Thermal-Hydraulics Analyses of the BR2 Maximum Nominal Heat Flux

    Energy Technology Data Exchange (ETDEWEB)

    Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Kim, Y. S. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hofman, G. L. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2015-02-01

    This memo describes the assumptions and references used in determining the thermal properties for the various materials used in the BR2 HEU (93% enriched in 235U) to LEU (19.75% enriched in 235U) conversion feasibility analysis. More specifically, this memo focuses on the materials contained within the pressure vessel (PV), i.e., the materials that are most relevant to the study of impact of the change of fuel from HEU to LEU. Section 2 provides a summary of the thermal properties in the form of tables while the following sections and appendices present the justification of these values. Section 3 presents a brief background on the approach used to evaluate the thermal properties of the dispersion fuel meat and specific heat capacity. Sections 4 to 7 discuss the material properties for the following materials: i) aluminum, ii) dispersion fuel meat (UAlx-Al and U-7Mo-Al), iii) beryllium, and iv) stainless steel. Section 8 discusses the impact of irradiation on material properties. Section 9 summarizes the material properties for typical operating temperatures. Appendix A elaborates on how to calculate dispersed phase’s volume fraction. Appendix B provides a revised methodology for determining the thermal conductivity as a function of burnup for HEU and LEU.

  2. 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

  3. Single well thermal tracer test, a new experimental set up for characterizing thermal transport in fractured media

    Science.gov (United States)

    de La Bernardie, Jérôme; Bour, Olivier; Guihéneuf, Nicolas; Chatton, Eliot; Labasque, Thierry; Longuevergne, Laurent; Le Lay, Hugo; Koch, Floriant; Gerard, Marie-Françoise; Le Borgne, Tanguy

    2017-04-01

    Thermal transport in fractured media depends on the hydrological properties of fractures and thermal characteristics of rock. Tracer tests using heat as tracer can thus be a good alternative to characterize fractured media for shallow geothermal needs. This study investigates the possibility of implementing a new thermal tracer test set up, the single well thermal tracer test, to characterize hydraulic and thermal transport properties of fractured crystalline rock. The experimental setup is based on injecting hot water in a fracture isolated by a double straddle packer in the borehole while pumping and monitoring the temperature in a fracture crossing the same borehole at greater elevation. One difficulty comes from the fact that injection and withdrawal are achieved in the same borehole involving thermal losses along the injection tube that may disturb the heat recovery signal. To be able to well localize the heat influx, we implemented a Fiber-Optic Distributed Temperature Sensing (FO-DTS) which allows the temperature monitoring with high spatial and temporal resolution (29 centimeters and 30 seconds respectively). Several tests, at different pumping and injection rates, were performed in a crystalline rock aquifer at the experimental site of Ploemeur (H+ observatory network). We show through signal processing how the thermal breakthrough may be extracted thanks to Fiber-Optic distributed temperature measurements. In particular, we demonstrate how detailed distributed temperature measurements were useful to identify different inflows and to estimate how much heat was transported and stored within the fractures network. Thermal breakthrough curves of single well thermal tracer tests were then interpreted with a simple analytical model to characterize hydraulic and thermal characteristics of the fractured media. We finally discuss the advantages of these tests compared to cross-borehole thermal tracer tests.

  4. IRIS thermal balance test within ESTEC LSS

    Science.gov (United States)

    Messidoro, Piero; Ballesio, Marino; Vessaz, J. P.

    1988-01-01

    The Italian Research Interim Stage (IRIS) thermal balance test was successfully performed in the ESTEC Large Space Simulator (LSS) to qualify the thermal design and to validate the thermal mathematical model. Characteristics of the test were the complexity of the set-up required to simulate the Shuttle cargo bay and allowing IRIS mechanism actioning and operation for the first time in the new LSS facility. Details of the test are presented, and test results for IRIS and the LSS facility are described.

  5. Development and industrial tests of the first LNG hydraulic turbine system in China

    Directory of Open Access Journals (Sweden)

    Jie Chen

    2016-10-01

    Full Text Available The cryogenic hydraulic turbine can be used to replace the conventional J–T valve for LNG or mixed refrigerant throttling and depressurization in a natural gas liquefaction plant. This advanced technology is not only to enhance the efficiency of the liquefaction plant, but to usher a new trend in the development of global liquefaction technologies. China has over 136 liquefaction plants, but the cryogenic hydraulic turbines have not been deployed in industrial utilization. In addition, these turbines cannot be manufactured domestically. In this circumstance, through working on the key technologies for LNG hydraulic turbine process & control system development, hydraulic model optimization design, structure design and manufacturing, the first domestic cryogenic hydraulic turbine with a flow rate of 40 m3/h was developed to recover the pressure energy from the LNG of cold box. The turbine was installed in the CNOOC Zhuhai Natural Gas Liquefaction Plant for industrial tests under multiple working conditions, including start-stop, variable flow rates and variable rotation speeds. Test results show that the domestic LNG cryogenic hydraulic turbine has satisfactory mechanical and operational performances at low temperatures as specified in design. In addition, the process & control system and frequency-conversion power-generation system of the turbine system are designed properly to automatically and smoothly replace the existing LNG J–T valve. As a result, the domestic LNG cryogenic hydraulic turbine system can improve LNG production by an average of 2% and generate power of 8.3 kW.

  6. 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.

  7. 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.

  8. Development of a thermal-hydraulic analysis code for annular fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vishnoi, A.K.; Chandraker, D.K.; Vijayan, P.K. [Bhabha Atomic Reseach Centre (BARC), Mumbai, Maharashtra (India)

    2012-03-15

    In this work a detailed study of the annular fuel has been carried out. A thermal hydraulics code, ANUFAN (Annular Fuel Analysis), based on the bundle average method, capable of modeling both internally and externally cooled annular fuel pins is developed. Code predictions have been compared with calculations from Korea Atomic Energy Research Institute (KAERI) and MIT. Heat transfer fraction difference between ANUFAN and RELAP was found about 1.7%. Analysis of a 54 - fuel rod assembly is carried out with 36 and 45 numbers of annular fuel pins keeping the same channel size and bundle power as of the solid fuel assembly. Fuel pin maximum temperature of the annular fuel is found much less than the solid fuel. MCHFR value for annular fuel is found much higher compared to that of the solid fuel of 54 - fuel rod assembly. The full paper covers the details of the computer code, the analysis carried out and the results obtained. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Fajardo, Laura Garcia, E-mail: laura.gf@cern.ch [European Organization for Nuclear Research (CERN), Geneva (Switzerland). Technology Department; Hernandez, Carlos Garcia; Mazaira, Leorlen Rojas, E-mail: cgh@instec.cu, E-mail: irojas@instec.cu [Higher Institute of Technologies and Applied Sciences (INSTEC), Habana (Cuba); Castells, Facundo Alberto Escriva, E-mail: aescriva@iqn.upv.es [University of Valencia (UV), Valencia (Spain). Energetic Engineering Institute; Lira, Carlos Brayner de Olivera, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (BRazil). Dept. de Engenharia Nuclear

    2013-07-01

    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)

  10. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2011-01-01

    Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

  11. 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)

  12. Thermal hydraulic core simulation of the MYRRHA Reactor in steady state operation

    Energy Technology Data Exchange (ETDEWEB)

    Ferandes, Gustavo H.N.; Ramos, Mário C.; Carvalho, Athos M.S.S.; Cabrera, Carlos E.V.; Costa, Antonella L.; Pereira, Claubia, E-mail: ghnfernandes@gmail.com, E-mail: marc5663@gmail.com, E-mail: athos1495@yahoo.com.br, E-mail: carlosvelcab@hotmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.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/CNPq (Brazil)

    2017-07-01

    MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) is a prototype nuclear subcritical reactor driven by a particle accelerator. As a special property, the reactor maintains the nuclear fission chain reaction by means of an external neutron source provided by a particle accelerator. The main aim of this work is to study two types of coolants, LBE (Lead-Bismuth Eutectic) and Na (Sodium) that are two strong candidates to be used in ADS systems as well as in Generation IV (GEN-IV) reactors. Firstly, it was developed a thermal hydraulic model of the MYRRHA core using the RELAP5-3D, considering LBE as coolant (original project). After this, the LBE was substituted by Na coolant to investigate the reactor behavior in such case. Results have demonstrated the high heat transfer capacity of the LBE coolant in this type of system. (author)

  13. Thermal-hydraulic processes involved in loss of residual heat removal during reduced inventory operation

    Energy Technology Data Exchange (ETDEWEB)

    Fletcher, C.D.; McHugh, P.R.; Naff, S.A.; Johnsen, G.W.

    1991-02-01

    This paper identifies the topics needed to understand pressurized water reactor response to an extended loss of residual heat removal event during refueling and maintenance outages. By identifying the possible plant conditions and cooling methods that would be used for each cooling mode, the controlling thermal-hydraulic processes and phenomena were identified. Controlling processes and phenomena include: gravity drain, core water boil-off, and reflux cooling processes. Important subcategories of the reflux cooling processes include: the initiation of reflux cooling from various plant conditions, the effects of air on reflux cooling, core level depression effects, issues regarding the steam generator secondaries, and the special case of boiler-condenser cooling with once-through steam generators. 25 refs., 6 figs., 1 tab.

  14. Thermal hydraulic codes for LWR safety analysis - present status and future perspective

    Energy Technology Data Exchange (ETDEWEB)

    Staedtke, H. [Commission of the European Union, Ispra (Italy)

    1997-07-01

    The aim of the present paper is to give a review on the current status and future perspective of present best-estimate Thermal Hydraulic codes. Reference is made to internationally well-established codes which have reached a certain state of maturity. The first part of the paper deals with the common basic code features with respect to the physical modelling and their numerical methods used to describe complex two-phase flow and heat transfer processes. The general predictive capabilities are summarized identifying some remaining code deficiencies and their underlying limitations. The second part discusses various areas including physical modelling, numerical techniques and informatic structure where the codes could be substantially improved.

  15. 2D Thermal Hydraulic Analysis and Benchmark in Support of HFIR LEU Conversion using COMSOL

    Energy Technology Data Exchange (ETDEWEB)

    Freels, James D [ORNL; Bodey, Isaac T [ORNL; Lowe, Kirk T [ORNL; Arimilli, Rao V [ORNL

    2010-09-01

    The research documented herein was funded by a research contract between the Research Reactors Division (RRD) of Oak Ridge National Laboratory (ORNL) and the University of Tennessee, Knoxville (UTK) Mechanical, Aerospace and Biomedical Engineering Department (MABE). The research was governed by a statement of work (SOW) which clearly defines nine specific tasks. This report is outlined to follow and document the results of each of these nine specific tasks. The primary goal of this phase of the research is to demonstrate, through verification and validation methods, that COMSOL is a viable simulation tool for thermal-hydraulic modeling of the High Flux Isotope Reactor (HFIR) core. A secondary goal of this two-dimensional phase of the research is to establish methodology and data base libraries that are also needed in the full three-dimensional COMSOL simulation to follow. COMSOL version 3.5a was used for all of the models presented throughout this report.

  16. Combined use of straddle packer testing and FLUTe profiling for hydraulic testing in fractured rock boreholes

    Science.gov (United States)

    Quinn, Patryk; Cherry, John A.; Parker, Beth L.

    2015-05-01

    A combination of high resolution hydraulic tests using straddle packers and transmissivity (T) profiling using the FLUTe flexible liner method (liner profiling) in densely fractured rock boreholes is shown to be efficient for the determination of the vertical distribution of T along the entire hole. The liner T profiling method takes a few hours or less to scan the entire borehole length resulting in a T profile. Under favorable conditions this method has good reliability for identifying the highest T zones identified by distinct decreases in liner velocity when these zones are covered by the descending liner. In contrast, for one short test interval (e.g., 1-2 m) the multiple-test, straddle-packer method takes a few hours to measure T with good precision and accuracy using a combination of steady-state and transient tests (e.g., constant head step tests, slug tests, and constant rate pumping tests). Because of the time consuming aspect of this multiple-test method, it is most efficient in each borehole to conduct straddle packer testing only in priority zones selected after assessment of other borehole data collected prior to packer testing. The T profile from the liner method is instrumental in selecting high permeable zones for application of the multiple-test method using straddle packers, which in turn, refines the T estimation from the liner profile. Results from three boreholes in densely fractured sandstone demonstrate this approach showing the synergistic use of the methods with emphasis on information important for determining hydraulic apertures.

  17. 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.

  18. Experiments and analytical studies related to blowdown and containment thermal hydraulics on CSF

    Energy Technology Data Exchange (ETDEWEB)

    Dutta, Anu, E-mail: adutta@barc.gov.in; Thangamani, I.; Shanware, V.M.; Rao, K.S.; Gera, B.; Ravi Kiran, A.; Goyal, P.; Verma, Vishnu; Sharma, P.K.; Agrawal, M.K.; Ganju, S.; Singh, R.K.

    2015-12-01

    Highlights: • Blowdown and containment thermal hydraulics experiments conducted in CSF. • RELAP5, ASTEC and CONTRAN codes used for analysis. • Containment peak pressure and temp predicted close to experimental values. • CONTRAN and ASTEC codes predict early containment depressurization. • Numerical procedure, benchmarked for loss of coolant accident in nuclear reactors. - Abstract: Containment Studies Facility (CSF) is volumetrically scaled down model of Indian Pressurized Heavy Water Reactor (IPHWR) containment for simulating LOCA/MSLB conditions which consists of concrete containment model (CM) and Primary Heat Transport Model (PHTM) vessel. Blowdown experiments at different initial vessel pressure conditions were recently conducted at CSF and the vessel and containment parameters such as pressure, temperature and level transients have been recorded during the experiments. The experimental results have been used for benchmarking of numerical procedure adopted for evaluating LOCA/MSLB conditions in nuclear containment. The numerical procedure involves simulation of blowdown phenomena using RELAP5 code for evaluating mass and energy discharge rates, which are then used for calculating containment pressure–temperature transients using ASTEC and in-house CONTRAN codes. Predictions of major parameters of vessel and containment model were found to be in good agreement with that of experimental data. In containment thermal hydraulic calculations, condensation heat transfer coefficient affects the containment pressure–temperature transients. Various empirical condensation models like Tagami, Uchida and Diffusion models have been incorporated in CONTRAN code and suitable condensation model has been identified for which predicted pressure values are close to the experimental one. The details of the experimental and analytical studies conducted are presented in this paper.

  19. ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) simulation of a loss of coolant accident in a space reactor

    Energy Technology Data Exchange (ETDEWEB)

    Roth, P.A.; Shumway, R.W.

    1988-01-01

    The Advanced Thermal Hydraulic Energy Network Analyzer (ATHENA) code was used to simulate a loss-of-coolant accident (LOCA) in a conceptual space reactor design. ATHENA provides the capability of simulating the thermal-hydraulic behavior of the wide variety of systems which are being considered for use in space reactors. Flow loops containing any one of several available working fluids may interact through thermal connections with other loops containing the same or a different working fluid. The code can be used to model special systems such as: heat pipes, point reactor kinetics, plant control systems, turbines, valves, and pumps. This work demonstrates the application of the thermal radiation model which has been recently incorporated into ATHENA and verifies the need for supplemental reactor cooling to prevent reactor fuel damage in the event of a LOCA.

  20. Current and anticipated use of thermal-hydraulic codes for BWR transient and accident analyses in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Arai, Kenji; Ebata, Shigeo [Toshiba Corp., Yokohama (Japan)

    1997-07-01

    This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding of the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.

  1. ENDF/B Thermal Data Testing

    CERN Document Server

    McCrosson, F J

    2001-01-01

    The thermal data testing group is concerned with establishing the merit of ENDF/B cross sections for the analysis of thermal systems. The integral experiments used in the testing are designed to analyze each of the phenomena identified in the familiar four-factor formula. For brevity, only the testing of the cross sections in uranium systems is described in this report.

  2. Thermal-hydraulic simulation of mercury target concepts for a pulsed spallation neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Siman-Tov, M.; Wendel, M.; Haines, J. [Oak Ridge National Lab., TN (United States)

    1996-06-01

    The Oak Ridge Spallation Neutron Source (ORSNS) is a high-power, accelerator-based pulsed spallation neutron source being designed by a multi-laboratory team led by Oak Ridge National Laboratory to achieve very high fluxes of neutrons for scientific experiments. The ORSNS is projected to have a 1 MW proton beam upgradable to 5 MW. About 60% of the beam power (1-5 MW, 17-83 kJ/pulse in 0.5 microsec at 60 cps) is deposited in the liquid metal (mercury) target having the dimensions of 65x30x10 cm (about 19.5 liter). Peak steady state power density is about 150 and 785 MW/m{sup 3} for 1 MW and 5 MW beam respectively, whereas peak pulsed power density is as high as 5.2 and 26.1 GW/m{sup 3}, respectively. The peak pulse temperature rise rate is 14 million C/s (for 5 MW beam) whereas the total pulse temperature rise is only 7 C. In addition to thermal shock and materials compatibility, key feasibility issues for the target are related to its thermal-hydraulic performance. This includes proper flow distribution, flow reversals, possible {open_quotes}hot spots{close_quotes} and the challenge of mitigating the effects of thermal shock through possible injection of helium bubbles throughout the mercury volume or other concepts. The general computational fluid dynamics (CFD) code CFDS-FLOW3D was used to simulate the thermal and flow distribution in three preliminary concepts of the mercury target. Very initial CFD simulation of He bubbles injection demonstrates some potential for simulating behavior of He bubbles in flowing mercury. Much study and development will be required to be able to `predict`, even in a crude way, such a complex phenomena. Future direction in both design and R&D is outlined.

  3. 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.

  4. The relevance of thermal hydraulics pipeline simulation as a regulatory support tool

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Patricia Mannarino; Santos, Almir Beserra dos [Agencia Nacional do Petroleo, Gas Natural e Biocombustiveis (ANP), Rio de Janeiro, RJ (Brazil)

    2009-07-01

    The capacity definition of a pipeline, along with its allocation, is very relevant to assure market transparency, nondiscriminatory access, security of supply, and also to give consistent signs for expansion needs. Nevertheless, the capacity definition is a controversial issue, and may widely vary depending on the technical and commercial assumptions made. To calculate a pipeline's nominal capacity, there are a variety of simulation tools, which include steady state, transient and on-line computer programs. It is desirable that the simulation tool is robust enough to predict the pipeline's capacity under different conditions. There are many variables that impact the flow through a pipeline, like gas characteristics, pipe and environmental variables. Designing a thermal model is a time-consuming task that requests understanding the level of detail need, in order to achieve success in its application. This article discusses the capacity definition, its role and calculation guidelines, describes ANP's experience with capacity calculation and further challenges according to the new regulation, and debates the role of thermal hydraulic simulation as a regulatory tool. (author)

  5. 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.

  6. A review on improving thermal-hydraulic performance of fin-and-tube heat exchangers

    Science.gov (United States)

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

    2017-10-01

    Fin-and-tube heat exchangers are one of the most common type of heat exchangers that are normally used in sectors that require small size and light weight but high heat transfer capabilities. Compact fin-and-tube heat exchangers experiences high convective thermal resistance at the air-side due to the thermo-physical properties of air. Thus, the purpose of this paper is to provide an overview of research works that are relevant to improving thermalhydraulic performance at the air-side of fin-and-tube heat exchangers. This paper covers a variety of parameters such as tube parameters like tube arrangement, tube shapes and tube inclination angles; extended surfaces such as different shapes of fins and different parameters of vortex generators like attack angles, shapes and locations. Overall, for most modifications there was increment in heat transfer but accompanied with a pressure drop penalty. However, this varies for different combinations of parameters thus this review is to help understand how every mentioned parameters influences the thermal-hydraulic performance.

  7. FEATURES OF RESOURCE TESTING OF THE HYDRAULIC BRAKE DRIVE ELEMENTS OF VEHICLES EQUIPPED WITH ABS

    Directory of Open Access Journals (Sweden)

    A. Revin

    2011-01-01

    Full Text Available The analysis of the resource testing facilities and methods of automobile brake cylinders in terms of ABS working process adequacy is carried out. A testing stand construction and a method of carrying out the resource testing of hydraulic drive elements of the automobile automated braking sys-tem is offered.

  8. Analysis of the Thermal and Hydraulic Stimulation Program at Raft River, Idaho

    Science.gov (United States)

    Bradford, Jacob; McLennan, John; Moore, Joseph; Podgorney, Robert; Plummer, Mitchell; Nash, Greg

    2017-05-01

    The Raft River geothermal field, located in southern Idaho, roughly 100 miles northwest of Salt Lake City, is the site of a Department of Energy Enhanced Geothermal System project designed to develop new techniques for enhancing the permeability of geothermal wells. RRG-9 ST1, the target stimulation well, was drilled to a measured depth of 5962 ft. and cased to 5551 ft. The open-hole section of the well penetrates Precambrian quartzite and quartz monzonite. The well encountered a temperature of 282 °F at its base. Thermal and hydraulic stimulation was initiated in June 2013. Several injection strategies have been employed. These strategies have included the continuous injection of water at temperatures ranging from 53 to 115 °F at wellhead pressures of approximately 275 psi and three short-term hydraulic stimulations at pressures up to approximately 1150 psi. Flow rates, wellhead and line pressures and fluid temperatures are measured continuously. These data are being utilized to assess the effectiveness of the stimulation program. As of August 2014, nearly 90 million gallons have been injected. A modified Hall plot has been used to characterize the relationships between the bottom-hole flowing pressure and the cumulative injection fluid volume. The data indicate that the skin factor is decreased, and/or the permeability around the wellbore has increased since the stimulation program was initiated. The injectivity index also indicates a positive improvement with values ranging from 0.15 gal/min psi in July 2013 to 1.73 gal/min psi in February 2015. Absolute flow rates have increased from approximately 20 to 475 gpm by February 2 2015. Geologic, downhole temperature and seismic data suggest the injected fluid enters a fracture zone at 5650 ft and then travels upward to a permeable horizon at the contact between the Precambrian rocks and the overlying Tertiary sedimentary and volcanic deposits. The reservoir simulation program FALCON developed at the Idaho National

  9. Preliminary fluid channel design and thermal-hydraulic analysis of glow discharge cleaning permanent electrode

    Energy Technology Data Exchange (ETDEWEB)

    Cai, Lijun, E-mail: cailj@swip.ac.cn [Southwestern Institute of Physics, Chengdu (China); Lin, Tao; Wang, Yingqiao; Wang, Mingxu [Southwestern Institute of Physics, Chengdu (China); Maruyama, So; Yang, Yu; Kiss, Gabor [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2016-11-01

    Highlights: • The plasma facing closure cap has to survive after 30,000 thermal heat load cycles. • 0.35 MW/m2 radiation heat load plus nuclear heat load are very challenging for stainless steel. • Multilayer structure has been designed by using advanced welding and drilling technology to solve the neutron heating problem. • Accurate volumetric load application in analysis model by CFX has been mastered. - Abstract: Glow discharge cleaning (GDC) shall be used on ITER device to reduce and control impurity and hydrogenic fuel out-gassing from in-vessel plasma facing components. After first plasma, permanent electrode (PE) will be used to replace Temporary Electrode (TE) for subsequent operation. Two fundamental scenarios i.e., GDC and Plasma Operation State (POS) should be considered for electrode design, which requires the heat load caused by plasma radiation and neutron heating must be taken away by cooling water flowing inside the electrode. In this paper, multilayer cooling channels inside PE are preliminarily designed, and snakelike route in each layer is adopted to improve the heat exchange. Detailed thermal-hydraulic analyses have been done to validate the design feasibility or rationality. The analysis results show that during GDC the cooling water inlet and outlet temperature difference is far less than the allowable temperature rise under water flow rate 0.15 kg/s compromised by many factors. For POS, the temperature rise and pressure drop are within the design goals, but high thermal stress occurs on the front surface of closure cap of electrode. After several iterations of optimization of the closure cap, the equivalent strain range after 30,000 loading cycles for POS is well below 0.3% design goals.

  10. Tethered satellite thermal design and test

    Science.gov (United States)

    Chapter, John J.

    1991-01-01

    The Tethered Satellite System (TSS) is the first Shuttle Orbiter mission that investigates electrodynamic phenomenon of a 20 km conductive tether, in space. The TSS Mission is planned for January 1992. The 'Deployer' that provides the mechanisms that control a tethered satellite is mounted on a Spacelab Pallet. The Deployer thermal design uses Multilayer Insulation (MLI), heaters, and the Spacelab payload freon loop. The pallet and Deployer are isolated from the space thermal environment with MLI that forms an enclosure that is a unique part of the thermal design. This paper describes the TSS thermal design, presents the analysis approach, and details the Deployer thermal balance test.

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

    Energy Technology Data Exchange (ETDEWEB)

    Anh Bui; Nam Dinh; Brian Williams

    2013-09-01

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

  12. Validation of coupled neutronic / thermal-hydraulic codes for VVER reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Mittag, S.; Grundmann, U.; Kliem, S.; Kozmenkov, Y.; Rindelhardt, U.; Rohde, U.; Weiss, F.-P.; Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K.-D.; Vanttola, T.; Haemaelaeinen, A.; Kaloinen, E.; Kereszturi, A.; Hegyi, G.; Panka, I.; Hadek, J.; Strmensky, C.; Darilek, P.; Petkov, P.; Stefanova, S.; Kuchin, A.; Khalimonchuk, V.; Hlbocky, P.; Sico, D.; Danilin, S.; Ionov, V.; Nikonov, S.; Powney, D.

    2004-08-01

    In recent years, the simulation methods for the safety analysis of nuclear power plants have been continuously improved to perform realistic calculations. Therefore in VALCO work package 2 (WP 2), the usual application of coupled neutron-kinetic / thermal-hydraulic codes to VVER has been supplemented by systematic uncertainty and sensitivity analyses. A comprehensive uncertainty analysis has been carried out. The GRS uncertainty and sensitivity method based on the statistical code package SUSA was applied to the two transients studied earlier in SRR-1/95: A load drop of one turbo-generator in Loviisa-1 (VVER-440), and a switch-off of one feed water pump in Balakovo-4 (VVER-1000). The main steps of these analyses and the results obtained by applying different coupled code systems (SMABRE - HEXTRAN, ATHLET - DYN3D, ATHLET - KIKO3D, ATHLET - BIPR-8) are described in this report. The application of this method is only based on variations of input parameter values. No internal code adjustments are needed. An essential result of the analysis using the GRS SUSA methodology is the identification of the input parameters, such as the secondary-circuit pressure, the control-assembly position (as a function of time), and the control-assembly efficiency, that most sensitively affect safety-relevant output parameters, like reactor power, coolant heat-up, and primary pressure. Uncertainty bands for these output parameters have been derived. The variation of potentially uncertain input parameter values as a consequence of uncertain knowledge can activate system actions causing quite different transient evolutions. This gives indications about possible plant conditions that might be reached from the initiating event assuming only small disturbances. In this way, the uncertainty and sensitivity analysis reveals the spectrum of possible transient evolutions. Deviations of SRR-1/95 coupled code calculations from measurements also led to the objective to separate neutron kinetics from

  13. Thermal hydraulic-Mechanic Integrated Simulation for Advanced Cladding Thermal Shock Fracture Analysis during Reflood Phase in LBLOCA

    Energy Technology Data Exchange (ETDEWEB)

    Son, Seong Min; Lee, You Ho; Cho, Jae Wan; Lee, Jeong Ik [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    This study suggested thermal hydraulic-mechanical integrated stress based methodology for analyzing the behavior of ATF type claddings by SiC-Duplex cladding LBLOCA simulation. Also, this paper showed that this methodology could predict real experimental result well. That concept for enhanced safety of LWR called Advanced Accident-Tolerance Fuel Cladding (ATF cladding, ATF) is researched actively. However, current nuclear fuel cladding design criteria for zircaloy cannot be apply to ATF directly because those criteria are mainly based on limiting their oxidation. So, the new methodology for ATF design criteria is necessary. In this study, stress based analysis methodology for ATF cladding design criteria is suggested. By simulating LBLOCA scenario of SiC cladding which is the one of the most promising candidate of ATF. Also we'll confirm our result briefly through comparing some facts from other experiments. This result is validating now. Some of results show good performance with 1-D failure analysis code for SiC fuel cladding that already developed and validated by Lee et al,. It will present in meeting. Furthermore, this simulation presented the possibility of understanding the behavior of cladding deeper. If designer can predict the dangerous region and the time precisely, it may be helpful for designing nuclear fuel cladding geometry and set safety criteria.

  14. Hydraulic testing of Salado Formation evaporites at the Waste Isolation Pilot Plant site: Second interpretive report

    Energy Technology Data Exchange (ETDEWEB)

    Beauheim, R.L. [Sandia National Labs., Albuquerque, NM (United States); Roberts, R.M.; Dale, T.F.; Fort, M.D.; Stensrud, W.A. [INTERA, Inc., Austin, TX (United States)

    1993-12-01

    Pressure-pulse, constant-pressure flow, and pressure-buildup tests have been performed in bedded evaporites of the Salado Formation at the Waste Isolation Pilot Plant (WIPP) site to evaluate the hydraulic properties controlling brine flow through the Salado. Transmissivities have been interpreted from six sequences of tests conducted on five stratigraphic intervals within 15 m of the WIPP underground excavations.

  15. Inverse modeling of hydraulic tests in fractured crystalline rock based on a transition probability geostatistical approach

    Science.gov (United States)

    Blessent, Daniela; Therrien, René; Lemieux, Jean-Michel

    2011-12-01

    This paper presents numerical simulations of a series of hydraulic interference tests conducted in crystalline bedrock at Olkiluoto (Finland), a potential site for the disposal of the Finnish high-level nuclear waste. The tests are in a block of crystalline bedrock of about 0.03 km3 that contains low-transmissivity fractures. Fracture density, orientation, and fracture transmissivity are estimated from Posiva Flow Log (PFL) measurements in boreholes drilled in the rock block. On the basis of those data, a geostatistical approach relying on a transitional probability and Markov chain models is used to define a conceptual model based on stochastic fractured rock facies. Four facies are defined, from sparsely fractured bedrock to highly fractured bedrock. Using this conceptual model, three-dimensional groundwater flow is then simulated to reproduce interference pumping tests in either open or packed-off boreholes. Hydraulic conductivities of the fracture facies are estimated through automatic calibration using either hydraulic heads or both hydraulic heads and PFL flow rates as targets for calibration. The latter option produces a narrower confidence interval for the calibrated hydraulic conductivities, therefore reducing the associated uncertainty and demonstrating the usefulness of the measured PFL flow rates. Furthermore, the stochastic facies conceptual model is a suitable alternative to discrete fracture network models to simulate fluid flow in fractured geological media.

  16. 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)

  17. Neutronics and thermal hydraulic analysis of TRIGA Mark II reactor using MCNPX and COOLOD-N2 computer code

    Science.gov (United States)

    Tiyapun, K.; Wetchagarun, S.

    2017-06-01

    The neutronic analysis of TRIGA Mark II reactor has been performed. A detailed model of the reactor core was conducted including standard fuel elements, fuel follower control rods, and irradiation devices. As the approach to safety nuclear design are based on determining the criticality (keff), reactivity worth, reactivity excess, hot rod power factor and power peaking of the reactor, the MCNPX code had been used to calculate the nuclear parameters for different core configuration designs. The thermal-hydraulic model has been developed using COOLOD-N2 for steady state, using the nuclear parameters and power distribution results from MCNPX calculation. The objective of the thermal-hydraulic model is to determine the thermal safety margin and to ensure that the fuel integrity is maintained during steady state as well as during abnormal condition at full power. The hot channel fuel centerline temperature, fuel surface temperature, cladding surface temperature, the departure from nucleate boiling (DNB) and DNB ratio were determined. The good agreement between experimental data and simulation concerning reactor criticality proves the reliability of the methodology of analysis from neutronic and thermal hydraulic perspective.

  18. Aespoe Hard Rock Laboratory. Backfill and Plug test. Hydraulic testing of core drilled boreholes in the ZEDEX drift

    Energy Technology Data Exchange (ETDEWEB)

    Ludvigson, Jan-Erik; Nordqvist, Rune; Ekman, Lennart; Hansson, Kent (GEOSIGMA AB, Uppsala (Sweden))

    2009-07-01

    The present report documents the performance and results of hydraulic testing in selected core boreholes in the Zedex drift. The holes will be used as rock instrumentation boreholes during the Backfill and Plug Test at Aespoe HRL. The testing involves both 1 m long boreholes with 56 mm diameter as well as longer boreholes c. 5 m, 8 m and 25 m long with 56 mm or 76 mm diameter. Only single-hole tests were performed. The tests were carried out as short-time constant head injection tests since all boreholes tested (except one) were non-flowing before tests. The injection phase was followed by a pressure recovery phase. Furthermore, the tests were carried out as single-packer tests. A specially designed test system was used for the tests. The main evaluation of the tests was performed on data from the recovery phase by a new approach based on a non-linear regression technique combined with a flow simulation model (SUTRA). The tests in the 1 m-holes (testing the interval c. 0.3-0.7 m in the rock perpendicular to the tunnel face) show that the hydraulic conductivity of the superficial rock around the Zedex drift in general is low. However, during testing in some boreholes, visible leakage in the rock occurred through superficial fractures into the tunnel. These fractures were mainly located in the floor of the Zedex drift and are probably blast-induced. These fractures have a high hydraulic conductivity. The tests in the longer boreholes show that the hydraulic conductivity further into the rock in general is below c. 1x10-10 m/s. Increased hydraulic conductivity (c.1.5x10-8 m/s) was only observed in the flowing borehole KXZSD8HL.

  19. A two-fluid two-phase model for thermal-hydraulic analysis of a U-tube steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Hung, Huanjen; Chieng, Chingchang; Pei, Baushei (National Tsinghua Univ, Hsinchu (Taiwan, Province of China)); Wang, Songfeng (Inst. of Nuclear Energy Research, Lungtan (Taiwan, Province of China))

    1993-02-01

    The Advanced Thermal-Hydraulic Analysis Code for Nuclear Steam Generators (ATHANS) was developed on the basis of the THERMIT-UTSG computer code for U-tube steam generators. The main features of the ATHANS model are as follows: (a) the equations are solved in cylindrical coordinates, (b) the number and the arrangement of the control volumes inside the steam generator can be chosen by the user, (c) the virtual mass effect is incorporated, and (d) the conjugate gradient squared method is employed to accelerate and improve the numerical convergence. The performance of the model is successfully validated by comparison with the test data from a Westinghouse model F steam generator at the Maanshan nuclear power plant. Better agreement with the test data can be obtained by a finer grid system using a cylindrical coordinate system and the virtual mass effect. With these advanced features, ATHANS provides the basic framework for further studies on the problems of steam generators, such as analyses of secondary-side corrosion and tube ruptures.

  20. ANTHEM2000{sup TM}: Integration of the ANTHEM Thermal Hydraulic Model in the ROSE{sup TM} Environment

    Energy Technology Data Exchange (ETDEWEB)

    Boire, R.; Nguyen, M; Salim, G. [CAE Electronics Ltd., Quebec (Canada)

    1999-07-01

    ROSEN{sup TM} is an object oriented, visual programming environment used for many applications, including the development of power plant simulators. ROSE provides an integrated suite of tools for the creation, calibration, test, integration, configuration management and documentation of process, electrical and I and C models. CAE recently undertook an ambitious project to integrate its two phase thermal hydraulic model ANTHEM{sup TM} into the ROSE environment. ANTHEM is a non equilibrium, non-homogenous model based on the drift flux formalism. CAE has used the model in numerous two phase applications for nuclear and fossil power plant simulators. The integration of ANTHEM into ROSE brings the full power of visual based programming to two phase modeling applications. Features include graphical model building, calibration tools, a superior test environment and process visualisation. In addition the integration of ANTHEM into ROSE makes it possible to easily apply the fidelity of ANTHEM to BOP applications. This paper describes the implementation of the ANTHEM model within the ROSE environment and gives examples of its use. (author)

  1. Experimental research in neutron physic and thermal-hydraulic at the CDTN Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Souza, Rose Mary G.P.; Ferreira, Andrea V.; Pinto, Antonio J.; Costa, Antonio C.L.; Rezende, Hugo C., E-mail: amir@cdtn.b, E-mail: souzarm@cdtn.b, E-mail: avf@cdtn.b, E-mail: ajp@cdtn.b, E-mail: aclc@cdtn.b, E-mail: hcr@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The IPR-R1 TRIGA (Training, Research, Isotopes production, General Atomics) at Nuclear Technology Development Center (CDTN) is a pool type reactor cooled by natural circulation of light water and an open surface. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world and characterized by inherent safety. The IPR-R1 is the only Brazilian nuclear research reactor available and able to perform experiments in which interaction between neutronic and thermal-hydraulic areas occurs. The IPR-R1 has started up on November 11th, 1960. At that time the maximum thermal power was 30 kW. The present forced cooling system was built in the 70th and the power was upgraded to 100 kW. Recently the core configuration and instrumentation was upgraded again to 250 kW at steady state, and is awaiting the license of CNEN to operate definitely at this new power. This paper describes the experimental research project carried out in the IPR-R1 reactor that has as objective evaluate the behaviour of the reactor operational parameters, and mainly to investigate the influence of temperature on the neutronic variables. The research was supported by Research Support Foundation of the State of Minas Gerais (FAPEMIG) and Brazilian Council for Scientific and Technological Development (CNPq). The research project meets the recommendations of the IAEA, for safety, modernization and development of strategic plan for research reactors utilization. This work is in line with the strategic objectives of Brazil, which aims to design and construct the Brazilian Multipurpose research Reactor (RMB). (author)

  2. Thermal Hydraulic Computational Fluid Dynamics Simulations and Experimental Investigation of Deformed Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Mays, Brian [AREVA Federal Services, Lynchburg, VA (United States); Jackson, R. Brian [TerraPower, Bellevue, WA (United States)

    2017-03-08

    The project, Toward a Longer Life Core: Thermal Hydraulic CFD Simulations and Experimental Investigation of Deformed Fuel Assemblies, DOE Project code DE-NE0008321, was a verification and validation project for flow and heat transfer through wire wrapped simulated liquid metal fuel assemblies that included both experiments and computational fluid dynamics simulations of those experiments. This project was a two year collaboration between AREVA, TerraPower, Argonne National Laboratory and Texas A&M University. Experiments were performed by AREVA and Texas A&M University. Numerical simulations of these experiments were performed by TerraPower and Argonne National Lab. Project management was performed by AREVA Federal Services. The first of a kind project resulted in the production of both local point temperature measurements and local flow mixing experiment data paired with numerical simulation benchmarking of the experiments. The project experiments included the largest wire-wrapped pin assembly Mass Index of Refraction (MIR) experiment in the world, the first known wire-wrapped assembly experiment with deformed duct geometries and the largest numerical simulations ever produced for wire-wrapped bundles.

  3. Study of thermal and hydraulic efficiency of supersonic tube of temperature stratification

    Science.gov (United States)

    Tsynaeva, Anna A.; Nikitin, Maxim N.; Tsynaeva, Ekaterina A.

    2017-10-01

    Efficiency of supersonic pipe for temperature stratification with finned subsonic surface of heat transfer is the major of this paper. Thermal and hydraulic analyses of this pipe were conducted to asses effects from installation of longitudinal rectangular and parabolic fins as well as studs of cylindrical, rectangular and parabolic profiles. The analysis was performed based on refined empirical equations of similarity, dedicated to heat transfer of high-speed gas flow with plain wall, and Kármán equation with Nikuradze constants. Results revealed cylindrical studs (with height-to-diameter ratio of 5:1) to be 1.5 times more efficient than rectangular fins of the same height. At the same time rectangular fins (with height-to-thickness ratio of 5:1) were tend to enhance heat transfer rate up to 2.67 times compared to bare walls from subsonic side of the pipe. Longitudinal parabolic fins have minuscule effect on combined efficiency of considered pipe since extra head losses void any gain of heat transfer. Obtained results provide perspective of increasing efficiency of supersonic tube for temperature stratification. This significantly broadens device applicability in thermostatting systems for equipment, cooling systems for energy converting machinery, turbine blades and aerotechnics.

  4. Theoretical investigation of the thermal hydraulic behaviour of a slab-type liquid metal target

    Energy Technology Data Exchange (ETDEWEB)

    Dury, T.V.; Smith, B.L. [Paul Scherrer Institut, Villigen (Switzerland)

    1996-06-01

    The thermal hydraulics codes CFDS-FLOW3D and ASTEC have been used to simulate a slabtype design of ESS spallation target. This design is single-skinned, and of tapering form (in the beam direction), with rounded sides in a cross-section through a plane normal to the beam. The coolant fluid used is mercury, under forced circulation, with an inlet temperature of 180{degrees}C. The goal of these computer studies was to understand the behaviour of the coolant flow, and hence to arrive at a design which optimises the heat extraction for a given beam power - in the sense of: (1) minimising the peak local fluid temperature within the target, (2) maintaining an acceptable temperature level and distribution over and through the target outer wall, (3) keeping the overall fluid pressure loss through the complete target to a minimum, (4) staying within the physical limits of overall size required, particularly in the region of primary spallation. Two- and three-dimensional models have been used, with different arrangements and design of internal baffles, and different coolant flow distributions at the target inlet. Nominal total inlet mass flow was 245 kg/s, and a heat deposition profile used which was based on the proton beam energy distribution. This gave a nominal total heat load of 3.23 MW - of which 8.2kW were deposited in the window steel.

  5. Thermal hydraulic investigations and optimization on the EVC system of a PWR by CFD simulation

    Energy Technology Data Exchange (ETDEWEB)

    Xi, Mengmeng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Tang, Mao [China Nuclear Power Design Engineering Co., Ltd., 518124 Shenzhen (China); Wang, Chenglong; Zheng, Meiyin; Qiu, Suizheng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China)

    2015-08-15

    Highlights: • This study constructs a full CFD model for the EVC system of a PWR. • The complex fluid and solid coupling is treated in the computation. • Primary characteristics of the velocity, pressure and temperature distributions in the EVC system are investigated. • The optimization of the EVC system with different inlet boundaries are performed. - Abstract: In order to optimize the design of Reactor Pit Ventilation (EVC) system in a Pressurized Water Reactor (PWR), it is necessary to study the characteristics of the velocity, pressure and temperature fields in the EVC system. A full computational fluid dynamics (CFD) model for the EVC system is constructed by a commercial CFD code, where the complex fluid and solid coupling is treated. The Shear Stress Transport (SST) model is adopted to perform the turbulence calculation. This paper numerically investigates the characteristics of the velocity, pressure and temperature distributions in the EVC system. In particular, the effects of inlet air parameters on the thermal hydraulic characteristics and the reactor pit structure are also discussed for the EVC system optimization. Simulations are carried out with different mesh sizes and boundary conditions for sensitivity analysis. The computational results are important references to optimize the design and verify the rationality of the EVC system.

  6. Stabilized FE simulation of prototype thermal-hydraulics problems with integrated adjoint-based capabilities

    Science.gov (United States)

    Shadid, J. N.; Smith, T. M.; Cyr, E. C.; Wildey, T. M.; Pawlowski, R. P.

    2016-09-01

    A critical aspect of applying modern computational solution methods to complex multiphysics systems of relevance to nuclear reactor modeling, is the assessment of the predictive capability of specific proposed mathematical models. In this respect the understanding of numerical error, the sensitivity of the solution to parameters associated with input data, boundary condition uncertainty, and mathematical models is critical. Additionally, the ability to evaluate and or approximate the model efficiently, to allow development of a reasonable level of statistical diagnostics of the mathematical model and the physical system, is of central importance. In this study we report on initial efforts to apply integrated adjoint-based computational analysis and automatic differentiation tools to begin to address these issues. The study is carried out in the context of a Reynolds averaged Navier-Stokes approximation to turbulent fluid flow and heat transfer using a particular spatial discretization based on implicit fully-coupled stabilized FE methods. Initial results are presented that show the promise of these computational techniques in the context of nuclear reactor relevant prototype thermal-hydraulics problems.

  7. Normal zone propagation and Thermal Hydraulic Quenchback in a cable-in-conduit superconductor

    Energy Technology Data Exchange (ETDEWEB)

    Lue, J.W.; Dresner, L.

    1993-11-01

    When a local normal zone appears in a cable-in-conduit superconductor, a slug of hot helium is produced. The pressure rises and the hot helium expands. Thus the normal zone propagation in such a conductor can be governed by the hot helium expansion, rather than the heat conduction along the conductor. The expansion of the hot helium compresses the cold helium outside of the normal zone. This raises th@ temperature of the cold helium. When the temperature rise reaches the current sharing limit, the superconductor in contact goes normal. Thus a rapid increase in normal zone propagation occur. This phenomenon is termed Thermal Hydraulic Quenchback (THQ). An experiment was performed to investigate this process. The existence of THQ was verified. Thresholds of THQ were also observed by varying the conductor current, the magnetic field, the temperature, and the initial normal zone length. When THQ occurred, normal zone propagation approaching the velocity of sound was observed. A better picture of THQ is obtained by a careful comparison of the data with analytical studies.

  8. Experimental and numerical thermal-hydraulics investigation of a molten salt reactor concept core

    Energy Technology Data Exchange (ETDEWEB)

    Yamaji, Bogdan; Aszodi, Attila [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2017-09-15

    In the paper measurement results of experimental modelling of a molten salt fast reactor concept will be presented and compared with three-dimensional computational fluid dynamics (CFD) simulation results. Purpose of this article is twofold, on one hand to introduce a geometry modification in order to avoid the disadvantages of the original geometry and discuss new measurement results. On the other hand to present an analysis in order to suggest a method of proper numerical modelling of the problem based on the comparison of calculation results and measurement data for the new, modified geometry. The investigated concept has a homogeneous cylindrical core without any internal structures. Previous measurements on the scaled and segmented plexiglas model of the concept core and simulation results have shown that this core geometry could be optimized for better thermal-hydraulics characteristics. In case of the original geometry strong undesired flow separation could develop, that could negatively affect the characteristics of the core from neutronics point of view as well. An internal flow distributor plate was designed and installed with the purpose of optimizing the flow field in the core by enhancing its uniformity. Particle image velocimetry (PIV) measurement results of the modified experimental model will be presented and compared to numerical simulation results with the purpose of CFD model validation.

  9. Thermal-Hydraulic Research Review and Cooperation Outcome for Light Water Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    In, Wang Kee; Shin, Chang Hwan; Lee, Chan; Chun, Tae Hyun; Oh, Dong Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Chi Young [Pukyong Nat’l Univ., Busan (Korea, Republic of)

    2016-12-15

    The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermalhydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermalhydraulic technology and the commercialization.

  10. THERMAL HYDRAULIC ISSUES OF CONTAINMENT FILTERED VENTING SYSTEM FOR A LONG OPERATING TIME

    Directory of Open Access Journals (Sweden)

    YOUNG SU NA

    2014-12-01

    Full Text Available This study investigated the thermal hydraulic issues in the Containment Filtered Venting System (CFVS for a long operating time using the MELCOR computer code. The modeling of the CFVS, including the models for pool scrubbing and the filter, was added to the input file for the OPR-1000, and a Station Blackout (SBO was chosen as an accident scenario. Although depressurization in the containment building as a primary objective of the CFVS was successful, the decontamination feature by scrubbing and filtering in the CFVS for a long operating time could fail by the continuous evaporation of the scrubbing solution. After the operation of the CFVS, the atmosphere temperature in the CFVS became slightly above the water saturation temperature owing to the release of an amount of steam with high temperature from the containment building to the scrubbing solution. Reduced pipe diameters at the inlet and outlet of the CFVS vessel mitigated the evaporation of scrubbing water by controlling the amount of high-temperature steam and the water saturation temperature.

  11. Summary of papers on current and anticipated uses of thermal-hydraulic codes

    Energy Technology Data Exchange (ETDEWEB)

    Caruso, R.

    1997-07-01

    The author reviews a range of recent papers which discuss possible uses and future development needs for thermal/hydraulic codes in the nuclear industry. From this review, eight common recommendations are extracted. They are: improve the user interface so that more people can use the code, so that models are easier and less expensive to prepare and maintain, and so that the results are scrutable; design the code so that it can easily be coupled to other codes, such as core physics, containment, fission product behaviour during severe accidents; improve the numerical methods to make the code more robust and especially faster running, particularly for low pressure transients; ensure that future code development includes assessment of code uncertainties as integral part of code verification and validation; provide extensive user guidelines or structure the code so that the `user effect` is minimized; include the capability to model multiple fluids (gas and liquid phase); design the code in a modular fashion so that new models can be added easily; provide the ability to include detailed or simplified component models; build on work previously done with other codes (RETRAN, RELAP, TRAC, CATHARE) and other code validation efforts (CSAU, CSNI SET and IET matrices).

  12. Current and anticipated uses of thermal-hydraulic codes in Spain

    Energy Technology Data Exchange (ETDEWEB)

    Pelayo, F.; Reventos, F. [Consejo de Seguridad Nuclear, Barcelona (Spain)

    1997-07-01

    Spanish activities in the field of Applied Thermal-Hydraulics are steadily increasing as the codes are becoming practicable enough to efficiently sustain engineering decision in the Nuclear Power industry. Before reaching this point, a lot of effort has been devoted to achieve this goal. This paper briefly describes this process, points at the current applications and draws conclusions on the limitations. Finally it establishes the applications where the use of T-H codes would be worth in the future, this in turn implies further development of the codes to widen the scope of application and improve the general performance. Due to the different uses of the codes, the applications mainly come from the authority, industry, universities and research institutions. The main conclusion derived from this paper establishes that further code development is justified if the following requisites are considered: (1) Safety relevance of scenarios not presently covered is established. (2) A substantial gain in margins or the capability to use realistic assumptions is obtained. (3) A general consensus on the licensability and methodology for application is reached. The role of Regulatory Body is stressed, as the most relevant outcome of the project may be related to the evolution of the licensing frame.

  13. Simplified thermal-hydraulic analysis of single phase natural circulation circuit with two heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Pinheiro, Larissa Cunha; Su, Jian, E-mail: larissa@lasme.coppe.ufrj.br, E-mail: sujian@lasme.coppe.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenhraria Nuclear; Cotta, Renato Machado, E-mail: cotta@mecanica.coppe.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (POLI/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Dept. de Engenharia Mecanica

    2015-07-01

    Single phase natural circulation circuits composed of two convective heat exchangers and connecting tubes are important for the passive heat removal from spent fuel pools (SFP). To keep the structural integrity of the stored spent fuel assemblies, continuously cooling has to be provided in order to avoid increase at the pool temperature and subsequent uncovering of the fuel and enhanced reaction between water and metal releasing hydrogen. Decay heat can achieve considerably high amounts of energy e.g. in the AP1000, considering the emergency fuel assemblies, the maximum heat decay will reach 13 MW in the 15th day (Westinghouse Electric Company, 2010). A highly efficient alternative to do so is by means of natural circulation, which is cost-effective compared to active cooling systems and is inherently safer since presents less associated devices and no external work is required. Many researchers have investigated safety and stability aspects of natural circulation loops (NCL). However, there is a lack of literature concerning the improvement of NCL through a standard unified methodology, especially for natural circulation circuits with two heat exchangers. In the present study, a simplified thermal-hydraulic analysis of single phase natural circulation circuit with two heat exchanges is presented. Relevant dimensionless key groups were proposed to for the design and safety analysis of a scaled NCL for the cooling of spent fuel storage pool with convective cooling and heating. (author)

  14. Investigation of Correlations for the Thermal-hydraulic Analysis of Liquid Metal Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Won Pyo; Jeong, Hae Yong; Lee, Yong Bum

    2007-08-15

    The present investigation is aimed at reducing favorable constitutive correlations from those developed for the thermal-hydraulic analysis of Liquid Metal Reactors (LMR), for reliable safety analyses of KALIMER. It is achieved by analyzing them in a point of their accuracies. The study is particularly important because its outcomes can provide an essential knowledge on their relative errors including their conservatisms to be analyzed in the future KALIMER licensing stage. The predictions of the correlations have been compared with available experimental data on both friction factors for the wired-wrapped rod bundles in the core and the heat transfer coefficients in the system. As a result, the heat transfer coefficient inside pipe currently featured in SSC-K has been found acceptable. It, however, has shown a discrepancy of about 60 % and thus an alternative one has been proposed for improvement. Meanwhile, the friction factor model in the current SSC-K has not shown a prominent discrepancy in prediction trend but it has not backed an enough theoretical basis so that another model has been proposed. A systematic assessment for effects of those factors to the conservatism must be fully understood for the future licensing stage, and systematic calculations must be followed by designing an assessment matrix. Besides, it is essential to conduct experiments under similar conditions for constitutive parts of geometries which represent the KALIMER design.

  15. Numerical simulation of thermal-hydraulic processes in the riser chamber of installation for clinker production

    Directory of Open Access Journals (Sweden)

    Borsuk Grzegorz

    2016-03-01

    Full Text Available Clinker burning process has a decisive influence on energy consumption and the cost of cement production. A new problem is to use the process of decarbonization of alternative fuels from waste. These issues are particularly important in the introduction of a two-stage combustion of fuel in a rotary kiln without the typical reactor-decarbonizator. This work presents results of numerical studies on thermal-hydraulic phenomena in the riser chamber, which will be designed to burn fuel in the system where combustion air is supplied separately from the clinker cooler. The mathematical model is based on a combination of two methods of motion description: Euler description for the gas phase and Lagrange description for particles. Heat transfer between particles of raw material and gas was added to the numerical calculations. The main aim of the research was finding the correct fractional distribution of particles. For assumed particle distribution on the first stage of work, authors noted that all particles were carried away by the upper outlet to the preheater tower, what is not corresponding to the results of experimental studies. The obtained results of calculations can be the basis for further optimization of the design and operating conditions in the riser chamber with the implementation of the system.

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

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Lee, Won Jae; Hwang, Moon Kyu; Lim, Hong Sik [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    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.The study was performed by reconsideration of the previous code assessment works and phenomena identification for essential accident scenario. Improvement areas of model development for auditing tool were identified based on the code comparison and PIRT results. Nine models have been improved significantly for the analysis of LOCA and Mon LOCA event. Conceptual problem or separate effect assessment have been performed to verify the model improvement. The linking calculation with CONTAIN 2.0 has been also enabled to establish the unified auditing code system. Analysis for the CANDU plant real transient and hypothetical LOCA bas been performed using the improved version. It has been concluded that the developed version can be utilized for the auditing analysis of LOCA and non-LOCA event for the CANDU reactor. 25 refs., 84 figs., 36 tabs. (Author)

  17. Are leaves 'freewheelin'? Testing for a wheeler-type effect in leaf xylem hydraulic decline.

    Science.gov (United States)

    Scoffoni, Christine; Sack, Lawren

    2015-03-01

    A recent study found that cutting shoots under water while xylem was under tension (which has been the standard protocol for the past few decades) could produce artefactual embolisms inside the xylem, overestimating hydraulic vulnerability relative to shoots cut under water after relaxing xylem tension (Wheeler et al. 2013). That study also raised the possibility that such a 'Wheeler effect' might occur in studies of leaf hydraulic vulnerability. We tested for such an effect for four species by applying a modified vacuum pump method to leaves with minor veins severed, to construct leaf xylem hydraulic vulnerability curves. We tested for an impact on leaf xylem hydraulic conductance (Kx ) of cutting the petiole and minor veins under water for dehydrated leaves with xylem under tension compared with dehydrated leaves after previously relaxing xylem tension. Our results showed no significant 'cutting artefact' for leaf xylem. The lack of an effect for leaves could not be explained by narrower or shorter xylem conduits, and may be due to lesser mechanical stress imposed when cutting leaf petioles, and/or to rapid refilling of emboli in petioles. These findings provide the first validation of previous measurements of leaf hydraulic vulnerability against this potential artefact. © 2014 John Wiley & Sons Ltd.

  18. Dependence of hydraulic conductivity estimates from slug tests on displacement depth

    Science.gov (United States)

    Sauter, M.; Brauchler, R.

    2009-12-01

    Slug tests are generally accepted as an efficient tool for the hydraulic characterisation of aquifers. They can readily be employed without major logistical efforts and allow in particular the characterisation of contaminated aquifers since there is no contaminated water that needs to be disposed of. They are especially suited for assessing low conductive rock formations, where pumping quickly has to be stopped because the wells dry up. In cases where a dense well network is available slug tests can provide detailed information about the spatial variability of hydraulic conductivity. Limitations are: a) the small volume of integration, b) problems of obtaining representative storage parameters, and c) a large number of effects, the data need to be corrected for, especially in highly conductive aquifers, e.g. inertia, skin effects etc.. During a characterisation study, performed in karstified rocks, it was observed that the hydraulic conductivities obtained from the slug test data depended on the depth of displacement, i.e. an increase in displacement from 2.5 m to 11 m resulted in a change of hydraulic conductivity by a factor of 10. This feature, which is attributed to the heterogeneous characteristics of the fracture / matrix system, has also been described by Streltsova (1988), who observed that whenever "the formation volume, influenced by the test is smaller than the representative formation volume required for fracture pattern replication". The following conceptual model explains this hydraulic behaviour. If the well is drilled within a compact matrix block, the hydraulic response of an aquifer section is mainly determined by the low matrix conductivity near the well. If however the well is connected to the fracture network, short tests will only excite the highly conductive fractures and the longer the duration of the test, the more the less conductive matrix will contribute to the test result. Therefore a systematic decrease in the hydraulic conductivity

  19. Exploratory use of periodic pumping tests for hydraulic characterization of faults

    Science.gov (United States)

    Cheng, Yan; Renner, Joerg

    2018-01-01

    Periodic pumping tests were conducted using a double-packer probe placed at four different depth levels in borehole GDP-1 at Grimselpass, Central Swiss Alps, penetrating a hydrothermally active fault. The tests had the general objective to explore the potential of periodic testing for hydraulic characterization of faults, representing inherently complex heterogeneous hydraulic features that pose problems for conventional approaches. Site selection reflects the specific question regarding the value of this test type for quality control of hydraulic stimulations of potential geothermal reservoirs. The performed evaluation of amplitude ratio and phase shift between pressure and flow rate in the pumping interval employed analytical solutions for various flow regimes. In addition to the previously presented 1-D and radial-flow models, we extended the one for radial flow in a system of concentric shells with varying hydraulic properties and newly developed one for bilinear flow. In addition to these injectivity analyses, we pursued a vertical-interference analysis resting on observed amplitude ratio and phase shift between the periodic pressure signals above or below packers and in the interval by numerical modeling of the non-radial-flow situation. When relying on the same model the order of magnitude of transmissivity values derived from the analyses of periodic tests agrees with that gained from conventional hydraulic tests. The field campaign confirmed several advantages of the periodic testing, for example, reduced constraints on testing time relative to conventional tests since a periodic signal can easily be separated from changing background pressure by detrending and Fourier transformation. The discrepancies between aspects of the results from the periodic tests and the predictions of the considered simplified models indicate a hydraulically complex subsurface at the drill site that exhibits also hydromechanical features in accord with structural information

  20. 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.

  1. Overview of the use of ATHENA for thermal-hydraulic analysis of systems with lead-bismuth coolant

    Energy Technology Data Exchange (ETDEWEB)

    C. B. Davis; A. S. Shieh

    2000-04-02

    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.

  2. Uncertainty and sensitivity analysis for the modeling of transients with interaction of thermal hydraulics and neutron kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Soeren Kliem; Siegfried Mittag [Forschungszentrum Rossendorf (FZR), Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Siegfried Langenbuch [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH, P.O.B. 13 28, D-85748 Garching (Germany)

    2005-07-01

    Full text of publication follows: The transition from the application of conservative models to the use of best-estimate models raises the question about the uncertainty of the obtained results. This question becomes especially important, if the best-estimate models should be used for safety analyses in the field of nuclear engineering. Different methodologies were developed to assess the uncertainty of the calculation results of computer simulation codes. One of them is the methodology developed by Gesellschaft fuer Anlagenund Reaktorsicherheit (GRS) which uses the statistical code package SUSA. In the past, this methodology was applied to the calculation results of the advanced thermal hydraulic system code ATHLET. In the frame of the recently finished EU FP5 funded research project VALCO, that methodology was extended and successfully applied to different coupled code systems, including the uncertainty analysis for neutronics. These code systems consist of a thermal hydraulic system code and a 3D neutron kinetic core model. One of the code systems applied was ATHLET coupled with the Rossendorf kinetics code DYN3D. Two real transients at NPPs with VVER-type reactors documented within the VALCO project were selected for analyses. One was the load drop of one of two turbines to house load level at the Loviisa-1 NPP (VVER-440), the second was a test with the switching-off of one of two main feed water pumps at the VVER-1000 Balakovo-4 NPP. The current paper is dedicated to the different steps of the use and implementation of the GRS methodology to coupled code systems and to the assessment of the results obtained by the DYN3D/ATHLET code. Based on the relevant physical processes in both transients, lists of possible sources of uncertainties were compiled. They are specific for the two transients. Besides control parameters like control rod movement and thermal hydraulic parameters like secondary side pressure, mass flow rates, pressurizer sprayer and heater

  3. Multi-scale, coupled Reactor Physics / Thermal-Hydraulics system and applications to the HPLWR 3 Pass Core

    OpenAIRE

    Monti, Lanfranco

    2009-01-01

    The HPLWR is an innovative reactor concept cooled with water at supercritical pressure. The pronounced changes of water properties with the heat-up demands advanced analyses tools which have been developed and successfully applied. Coupled neutronic/thermal-hydraulic analyses have been performed for the whole core and the coupled solution has been successively investigated at sub-channel resolution evaluating local quantities. The obtained results represent a new quality in core analyses.

  4. A comparative assessment of independent thermal-hydraulic models for research reactors: The RSG-GAS case

    Energy Technology Data Exchange (ETDEWEB)

    Chatzidakis, S., E-mail: schatzid@purdue.edu [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47907 (United States); Hainoun, A. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400, San Carlos de Bariloche, Rio Negro (Argentina); Alhabet, F. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Francioni, F. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400, San Carlos de Bariloche, Rio Negro (Argentina); Ikonomopoulos, A. [Institute of Nuclear and Radiological Sciences, Energy, Technology and Safety, National Center for Scientific Research ‘Demokritos’, 15130, Aghia Paraskevi, Athens (Greece); Ridikas, D. [Department of Nuclear Sciences and Applications, International Atomic Energy Agency, Vienna International Centre, A-1400 Vienna (Austria)

    2014-03-15

    Highlights: • Increased use of thermal-hydraulic codes requires assessment of important phenomena in RRs. • Three independent modeling teams performed analysis of loss of flow transient. • Purpose of this work is to examine the thermal-hydraulic codes response. • To perform benchmark analysis comparing the different codes with experimental measurements. • To identify the impact of the user effect on the computed results, performed with the same codes. - Abstract: This study presents the comparative assessment of three thermal-hydraulic codes employed to model the Indonesian research reactor (RSG-GAS) and simulate the reactor behavior under steady state and loss of flow transient (LOFT). The RELAP5/MOD3, MERSAT and PARET-ANL thermal-hydraulic codes are used by independent research groups to perform benchmark analysis against measurements of coolant and clad temperatures, conducted on an instrumented fuel element inside RSG-GAS core. The results obtained confirm the applicability of RELAP5/MOD3, MERSAT and PARET-ANL on the modeling of loss of flow transient in research reactors. In particular, the three codes are able to simulate flow reversal from downward forced to upward natural convection after pump trip and successful reactor scram. The benchmark results show that the codes predict maximum clad temperature of hot channel conservatively with a maximum overestimation of 27% for RELAP5/MOD3, 17% for MERSAT and 8% for PARET-ANL. As an additional effort, the impact of user effect on the simulation results has been assessed for the code RELAP5/MOD3, where the main differences among the models are presented and discussed.

  5. Development of Effective Algorithm for Coupled Thermal-Hydraulics – Neutron-Kinetics Analysis of Reactivity Transient

    OpenAIRE

    Peltonen, Joanna

    2009-01-01

    Analyses of nuclear reactor safety have increasingly required coupling of full three dimensional neutron kinetics (NK) core models with system transient thermal-hydraulics (TH) codes. To produce results within a reasonable computing time, the coupled codes use different spatial description of the reactor core. The TH code uses few, typically 5 to 20 TH channels, which represent the core. The NK code uses explicit node for each fuel assembly. Therefore, a spatial mapping of coarse grid TH and ...

  6. Estimating hydraulic properties of volcanic aquifers using constant-rate and variable-rate aquifer tests

    Science.gov (United States)

    Rotzoll, K.; El-Kadi, A. I.; Gingerich, S.B.

    2007-01-01

    In recent years the ground-water demand of the population of the island of Maui, Hawaii, has significantly increased. To ensure prudent management of the ground-water resources, an improved understanding of ground-water flow systems is needed. At present, large-scale estimations of aquifer properties are lacking for Maui. Seven analytical methods using constant-rate and variable-rate withdrawals for single wells provide an estimate of hydraulic conductivity and transmissivity for 103 wells in central Maui. Methods based on constant-rate tests, although not widely used on Maui, offer reasonable estimates. Step-drawdown tests, which are more abundantly used than other tests, provide similar estimates as constant-rate tests. A numerical model validates the suitability of analytical solutions for step-drawdown tests and additionally provides an estimate of storage parameters. The results show that hydraulic conductivity is log-normally distributed and that for dike-free volcanic rocks it ranges over several orders of magnitude from 1 to 2,500 m/d. The arithmetic mean, geometric mean, and median values of hydraulic conductivity are respectively 520, 280, and 370 m/d for basalt and 80, 50, and 30 m/d for sediment. A geostatistical approach using ordinary kriging yields a prediction of hydraulic conductivity on a larger scale. Overall, the results are in agreement with values published for other Hawaiian islands. ?? 2007 American Water Resources Association.

  7. Methods of Testing Thermal Insulation and Associated Test Apparatus

    Science.gov (United States)

    2004-01-01

    The system and method for testing thermal insulation uses a cryostatic insulation tester having a vacuum chamber and a cold mass including a test chamber and upper and lower guard chambers adjacent thereto. The thermal insulation is positioned within the vacuum chamber and adjacent the cold mass. Cryogenic liquid is supplied to the test chamber, upper guard and lower guard to create a first gas layer in an upper portion of the lower guard chamber and a second gas layer in an upper portion of the test chamber. Temperature are sensed within the vacuum chamber to test the thermal insulation.

  8. Stochastic joint inversion of hydrogeophysical data for salt tracer test monitoring and hydraulic conductivity imaging

    Science.gov (United States)

    Jardani, A.; Revil, A.; Dupont, J. P.

    2013-02-01

    The assessment of hydraulic conductivity of heterogeneous aquifers is a difficult task using traditional hydrogeological methods (e.g., steady state or transient pumping tests) due to their low spatial resolution. Geophysical measurements performed at the ground surface and in boreholes provide additional information for increasing the resolution and accuracy of the inverted hydraulic conductivity field. We used a stochastic joint inversion of Direct Current (DC) resistivity and self-potential (SP) data plus in situ measurement of the salinity in a downstream well during a synthetic salt tracer experiment to reconstruct the hydraulic conductivity field between two wells. The pilot point parameterization was used to avoid over-parameterization of the inverse problem. Bounds on the model parameters were used to promote a consistent Markov chain Monte Carlo sampling of the model parameters. To evaluate the effectiveness of the joint inversion process, we compared eight cases in which the geophysical data are coupled or not to the in situ sampling of the salinity to map the hydraulic conductivity. We first tested the effectiveness of the inversion of each type of data alone (concentration sampling, self-potential, and DC resistivity), and then we combined the data two by two. We finally combined all the data together to show the value of each type of geophysical data in the joint inversion process because of their different sensitivity map. We also investigated a case in which the data were contaminated with noise and the variogram unknown and inverted stochastically. The results of the inversion revealed that incorporating the self-potential data improves the estimate of hydraulic conductivity field especially when the self-potential data were combined to the salt concentration measurement in the second well or to the time-lapse cross-well electrical resistivity data. Various tests were also performed to quantify the uncertainty in the inverted hydraulic conductivity

  9. Parametric studies by means of uncertainty and sensitivity methods for coupled thermal-hydraulic/neutron-physics application

    Energy Technology Data Exchange (ETDEWEB)

    Jaeger, W.; Sanchez, V.; Cheng, X. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Neutron Physics and Reactor Technology; Monti, L. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Nuclear and Energy Technologies; Hurtado, A. [Technical Univ. of Dresden (Germany). Inst. of Power Engineering

    2011-07-01

    At the Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT), the development and validation of coupled codes systems is one major activity. In this paper, a 2-step method is proposed to perform uncertainty and sensitivity analysis of a nuclear fuel bundle. At first, the SUSA package (Software system for Uncertainty and Sensitivity Analysis), 2 is applied to the thermal hydraulic results of the TRACE (TRACE/RELAP Advanced Computational Engine) code to identify crucial thermal hydraulic parameter combinations which are successively used in the TH/NP coupled system TRACEERANOS to account for the neutronic feedbacks. This 2-step method was applied since the TRACE-ERANOS system runs 1 input in approximately 1 day (depending on the computer configurations). Since the uncertainty and sensitivity analysis requires about 100 runs of the thermal hydraulic input (with altered parameters, running within minutes) an integral TRACE-SUSA-ERANOS analysis would need around 100 days. For this analysis a fuel assembly model of the HPLWR (High Performance Light Water Reactor) was selected. Due to the general structure of the coupling and code communication scripts, the system can be used for any kind of reactor/system which can be described with TRACE and ERANOS (e.g., fast systems) while SUSA can be applied to anything. (orig.)

  10. Digital Measuring Devices Used for Injector Hydraulic Test

    Directory of Open Access Journals (Sweden)

    S. N. Leontiev

    2015-01-01

    Full Text Available To ensure a high specific impulse of the LRE (liquid-propellant engine chamber it is necessary to have optimally organized combustion of the fuel components. This can be ensured by choosing the optimum geometry of gas-dynamic contour of the LRE combustor, as well as by improving the sputtering processes and mixing the fuel components, for example, by selection of the optimum type, characteristics, and location of injectors on the mixing unit of the chamber.These particular reasons arise the interest in the injector characteristics in terms of science, and technological aspects determine the need for control of underlying design parameters in their manufacture.The objective of this work is to give an experimental justification on used digital measurement instrumentation and research the hydraulic characteristics of injectors.To determine injector parameters most widely were used the units with sectional collectors. A technique to control injector parameters using the sectional collectors involves spraying the liquid by injector at a given pressure drop on it for a certain time (the longer, the higher the accuracy and reliability of the results and then determining the amount of liquid in each section to calculate the required parameters of injector.In this work the liquid flow through the injector was determined by high-precision flowmeters FLONET FN2024.1 of electromagnetic type, which have very high metrological characteristics, in particular a flow rate error does not exceed 0.5% in a range of water flow from Qmin= 0.0028 l/s to Qmax Qmax = 0.28 l/s. To determine the coefficient of uneven spray were used differential pressure sensors DMD 331-ASLX of company "DB Sensors RUS", which have an error of 0.075% with a range of differential pressure 0 ... 5 kPa. Measuring complex MIC-200 of company "NPP Measure" and WinPos software for processing array information provided entry, recording, and processing of all the data of the experiment.In this

  11. Final report of the 'Nordic thermal-hydraulic and safety network (NOTNET)' - Project

    Energy Technology Data Exchange (ETDEWEB)

    Tuunanen, J.; Tuomainen, M. [VTT Processes (Finland)

    2005-04-01

    A Nordic network for thermal-hydraulics and nuclear safety research was started. The idea of the network is to combine the resources of different research teams in order to carry out more ambitious and extensive research programs than would be possible for the individual teams. From the very beginning, the end users of the research results have been integrated to the network. Aim of the network is to benefit the partners involved in nuclear energy in the Nordic Countries (power companies, reactor vendors, safety regulators, research units). First task within the project was to describe the resources (personnel, know-how, simulation tools, test facilities) of the various teams. Next step was to discuss with the end users about their research needs. Based on these steps, few most important research topics with defined goals were selected, and coarse road maps were prepared for reaching the targets. These road maps will be used as a starting point for planning the actual research projects in the future. The organisation and work plan for the network were established. National coordinators were appointed, as well as contact persons in each participating organisation, whether research unit or end user. This organisation scheme is valid for the short-term operation of NOTNET when only Nordic organisations take part in the work. Later on, it is possible to enlarge the network e.g. within EC framework programme. The network can now start preparing project proposals and searching funding for the first common research projects. (au)

  12. Thermal control wall prototype and test results

    Energy Technology Data Exchange (ETDEWEB)

    Nakao, M.; Ohshima, K.; Jitsukawa, H.

    1986-01-01

    This paper describes a heat exchanger prototype and test results. The heat exchanger, called a thermal control wall, functions as a skin wall and as a means to vary the exterior wall thermal resistance of a building. Test results confirm that the capacity of the TCW is influenced by solar radiation. Furthermore, this TCW capacity can be evaluated by an overall heat transmission coefficient defined using the same sol air temperature difference as for a conventional wall.

  13. Development of CFD thermal hydraulics and neutron kinetics coupling methodologies for the prediction of local safety parameters for light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Perez Manes, Jorge

    2013-02-26

    This dissertation contributes to the development of high-fidelity coupled neutron kinetic and thermal hydraulic simulation tools with high resolution of the spatial discretization of the involved domains for the analysis of Light Water Reactors transient scenarios.

  14. Efficiency testing of hydraulic artificial muscles with variable recruitment using a linear dynamometer

    Science.gov (United States)

    Chipka, Jordan B.; Meller, Michael A.; Garcia, Ephrahim

    2015-03-01

    When a task calls for consistent, large amounts of power output, hydraulic actuation is a popular choice. However, for certain systems that require short bursts of high power, followed by a period of low power, the inefficiencies of hydraulics become apparent. One system that fits this description is a legged robot. McKibben muscles prove to be a wise choice for use on legged robots due to their light weight, high force capability, and inherent compliance. Variable recruitment, another novel concept for hydraulic actuation, offers the ability to further improve efficiency for hydraulic systems. This paper will discuss the efficiency characterization of variable recruitment McKibben muscles intended for use on a bipedal robot, but will focus on the novel test apparatus to do so. This device is a hydraulic linear dynamometer that will be controlled such that the muscles experience similar force-stroke levels to what will be required on a bipedal robot. The position of the dynamometer's drive cylinder will be controlled so that the muscles experience the proper position trajectory that will be needed on the robot. The pressure of the muscles will be controlled such that the force they experience will mimic the forces that occur on the robot while walking. Hence, these dynamic tests will ensure that the muscle bundles will meet the force-stroke requirements for the given robot. Once these muscle bundles are integrated onto the walking robot, the power savings of variable recruitment McKibben muscle bundles compared to the traditional hydraulic system will be demonstrated.

  15. Simulations of thermal-hydraulic processes in heat exchangers- station of the cogeneration power plant

    Energy Technology Data Exchange (ETDEWEB)

    Studovic, M.; Stevanovic, V.; Ilic, M.; Nedeljkovic, S. [Faculty of Mechanical Engineering of Belgrade (Croatia)

    1995-12-31

    Design of the long district heating system to Belgrade (base load 580 MJ/s) from Thermal Power Station `Nikola Tesla A`, 30 km southwest from the present gas/oil burning boilers in New Belgrade, is being conducted. The mathematical model and computer code named TRP are developed for the prediction of the design basis parameters of heat exchangers station, as well as for selection of protection devices and formulation of operating procedures. Numerical simulations of heat exchangers station are performed for various transient conditions: up-set and abnormal. Physical model of multi-pass, shell and tube heat exchanger in the station represented is by unique steam volume, and with space discretised nodes both for water volume and tube walls. Heat transfer regimes on steam and water side, as well as hydraulic calculation were performed in accordance with TEMA standards for transient conditions on both sides, and for each node on water side. Mathematical model is based on balance equations: mass and energy for lumped parameters on steam side, and energy balances for tube walls and water in each node. Water mass balance is taken as boundary/initial condition or as specified control function. The physical model is proposed for (s) heat exchangers in the station and (n) water and wall volumes. Therefore, the mathematical model consists of 2ns+2, non-linear differential equations, including equations of state for water, steam and tube material, and constitutive equations for heat transfer on steam and water side, solved by the Runge-Kutt method. Five scenarios of heat exchangers station behavior have been simulated with the TRP code and obtained results are presented. (author)

  16. Simulating High Flux Isotope Reactor Core Thermal-Hydraulics via Interdimensional Model Coupling

    Energy Technology Data Exchange (ETDEWEB)

    Travis, Adam R [ORNL

    2014-05-01

    A coupled interdimensional model is presented for the simulation of the thermal-hydraulic characteristics of the High Flux Isotope Reactor core at Oak Ridge National Laboratory. The model consists of two domains a solid involute fuel plate and the surrounding liquid coolant channel. The fuel plate is modeled explicitly in three-dimensions. The coolant channel is approximated as a twodimensional slice oriented perpendicular to the fuel plate s surface. The two dimensionally-inconsistent domains are linked to one another via interdimensional model coupling mechanisms. The coupled model is presented as a simplified alternative to a fully explicit, fully three-dimensional model. Involute geometries were constructed in SolidWorks. Derivations of the involute construction equations are presented. Geometries were then imported into COMSOL Multiphysics for simulation and modeling. Both models are described in detail so as to highlight their respective attributes in the 3D model, the pursuit of an accurate, reliable, and complete solution; in the coupled model, the intent to simplify the modeling domain as much as possible without affecting significant alterations to the solution. The coupled model was created with the goal of permitting larger portions of the reactor core to be modeled at once without a significant sacrifice to solution integrity. As such, particular care is given to validating incorporated model simplifications. To the greatest extent possible, the decrease in solution time as well as computational cost are quantified versus the effects such gains have on the solution quality. A variant of the coupled model which sufficiently balances these three solution characteristics is presented alongside the more comprehensive 3D model for comparison and validation.

  17. Parametric studies on containment thermal hydraulic loads during high pressure melt ejection in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Silde, A.; Lindholm, I. [VTT Energy, Espoo (Finland)

    1997-12-01

    The containment thermal hydraulic loads during high pressure melt ejection in a Nordic BWR are studied parametrically with the CONTAIN and the MELCOR codes. The work is part of the Nordic RAK-2 project. The containment analyses were divided into two categories according to composition of the discharged debris: metallic and oxidic debris cases. In the base case with highly metallic debris, all sources from the reactor coolant system to the containment were based on the MELCOR/BH calculation. In the base case with the oxidic debris, the source data was specified assuming that {approx} 15% of the whole core material inventory and 34,000 kg of saturated water was discharged from the reactor pressure vessel (RPV) during 30 seconds. In this case, the debris consisted mostly of oxides. The highest predicted containment pressure peaks were about 8.5 bar. In the scenarios with highly metallic debris source, very high gas temperature of about 1900 K was predicted in the pedestal, and about 1400 K in the upper drywell. The calculations with metallic debris were sensititive to model parameters, like the particle size and the parameters, which control the chemical reaction kinetics. In the scenarios with oxidic debris source, the predicted pressure peaks were comparable to the cases with the metallic debris source. The maximum gas temperatures (about 450-500 K) in the containment were, however, significantly lower than in the respective metallic debris case. The temperatures were also insensitive to parametric variations. In addition, one analysis was performed with the MELCOR code for benchmarking of the MELCOR capabilities against the more detailed CONTAIN code. The calculations showed that leak tightness of the containment penetrations could be jeopardized due to high temperature loads, if a high pressure melt ejection occurred during a severe accident. Another consequence would be an early containment venting. (au). 28 refs.

  18. ANTEO+: A subchannel code for thermal-hydraulic analysis of liquid metal cooled systems

    Energy Technology Data Exchange (ETDEWEB)

    Lodi, F., E-mail: francesco.lodi5@unibo.it [DIN – Laboratory of Montecuccolino, University of Bologna, Via dei Colli 16, 40136 Bologna (Italy); Grasso, G., E-mail: giacomo.grasso@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Mattioli, D., E-mail: davide.mattioli@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Sumini, M., E-mail: marco.sumini@unibo.it [DIN – Laboratory of Montecuccolino, University of Bologna, Via dei Colli 16, 40136 Bologna (Italy)

    2016-05-15

    Highlights: • The code structure is presented in detail. • The performed validation is outlined. • Results are critically discussed assessing code accuracy. • Conclusions are drawn and ground for future work identified. - Abstract: Liquid metal cooled fast reactors are promising options for achieving the high degrees of safety and sustainability demanded by the Generation IV paradigm. Among the critical aspects to be addressed in the design process, thermal-hydraulics is one of the most challenging; in order to embed safety in the core conceptualization, these aspects are to be considered at the very beginning of the design process, and translated in a design perspective. For achieving these objectives the subchannel code ANTEO+ has been conceived, able to simulate pin bundle arrangements cooled by liquid metals. The main purposes of ANTEO+ are simplifying the problem description maintaining the required accuracy, enabling a more transparent interface with the user, and having a clear and identifiable application domain, in order to help the user interpreting the results and, mostly, defining their confidence. Since ANTEO+ relies on empirical correlations, the validation phase is of paramount importance along with a clear discussion on the simplifications adopted in modeling the conservation equations. In the present work a detailed description of ANTEO+ structure is given along with a thorough validation of the main models implemented for flow split, pressure drops and subchannel temperatures. The analysis confirmed the ability of ANTEO+ in reproducing experimental data in its anticipated validity domain, with a relatively high degree of accuracy when compared to other classical subchannel tools like ENERGY-II, COBRA-IV-I-MIT and BRS-TVS.

  19. Development of a multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3 and its verification

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    A multi-dimensional realistic thermal-hydraulic system analysis code, MARS version 1.3 has been developed. Main purpose of MARS 1.3 development is to have the realistic analysis capability of transient two-phase thermal-hydraulics of Pressurized Water Reactors (PWRs) especially during Large Break Loss of Coolant Accidents (LBLOCAs) where the multi-dimensional phenomena domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, three-dimensional (3D) reactor vessel analysis code, and RELAP5/MOD3.2.1.2, one-dimensional (1D) reactor system analysis code., Developmental requirements for MARS are chosen not only to best utilize the existing capability of the codes but also to have the enhanced capability in code maintenance, user accessibility, user friendliness, code portability, code readability, and code flexibility. For the maintenance of existing codes capability and the enhancement of code maintenance capability, user accessibility and user friendliness, MARS has been unified to be a single code consisting of 1D module (RELAP5) and 3D module (COBRA-TF). This is realized by implicitly integrating the system pressure matrix equations of hydrodynamic models and solving them simultaneously, by modifying the 1D/3D calculation sequence operable under a single Central Processor Unit (CPU) and by unifying the input structure and the light water property routines of both modules. In addition, the code structure of 1D module is completely restructured using the modular data structure of standard FORTRAN 90, which greatly improves the code maintenance capability, readability and portability. For the code flexibility, a dynamic memory management scheme is applied in both modules. MARS 1.3 now runs on PC/Windows and HP/UNIX platforms having a single CPU, and users have the options to select the 3D module to model the 3D thermal-hydraulics in the reactor vessel or other

  20. Non-destructive testing principles and accurate evaluation of the hydraulic measure impact range using the DC method

    Science.gov (United States)

    Qiu, Liming; Shen, Rongxi; Song, Dazhao; Wang, Enyuan; Liu, Zhentang; Niu, Yue; Jia, Haishan; Xia, Shankui; Zheng, Xiangxin

    2017-12-01

    An accurate and non-destructive evaluation method for the hydraulic measure impact range in coal seams is urgently needed. Aiming at the application demands, a theoretical study and field test are presented using the direct current (DC) method to evaluate the impact range of coal seam hydraulic measures. We firstly analyzed the law of the apparent resistivity response of an abnormal conductive zone in a coal seam, and then investigated the principle of non-destructive testing of the coal seam hydraulic measure impact range using the DC method, and used an accurate evaluation method based on the apparent resistivity cloud chart. Finally, taking hydraulic fracturing and hydraulic flushing as examples, field experiments were carried out in coal mines to evaluate the impact ranges. The results showed that: (1) in the process of hydraulic fracturing, coal conductivity was enhanced by high-pressure water in the coal seam, and after hydraulic fracturing, the boundary of the apparent resistivity decrease area was the boundary impact range. (2) In the process of hydraulic flushing, coal conductivity was reduced by holes and cracks in the coal seam, and after hydraulic flushing, the boundary of the apparent resistivity increase area was the boundary impact range. (3) After the implementation of the hydraulic measures, there may be some blind zones in the coal seam; in hydraulic fracturing blind zones, the apparent resistivity increased or stayed constant, while in hydraulic flushing blind zones, the apparent resistivity decreased or stayed constant. The DC method realized a comprehensive and non-destructive evaluation of the impact range of the hydraulic measures, and greatly reduced the time and cost of evaluation.

  1. Characterizing hydraulic fractures in shale gas reservoirs using transient pressure tests

    Directory of Open Access Journals (Sweden)

    Cong Wang

    2015-06-01

    This work presents an unconventional gas reservoir simulator and its application to quantify hydraulic fractures in shale gas reservoirs using transient pressure data. The numerical model incorporates most known physical processes for gas production from unconventional reservoirs, including two-phase flow of liquid and gas, Klinkenberg effect, non-Darcy flow, and nonlinear adsorption. In addition, the model is able to handle various types and scales of fractures or heterogeneity using continuum, discrete or hybrid modeling approaches under different well production conditions of varying rate or pressure. Our modeling studies indicate that the most sensitive parameter of hydraulic fractures to early transient gas flow through extremely low permeability rock is actually the fracture-matrix contacting area, generated by fracturing stimulation. Based on this observation, it is possible to use transient pressure testing data to estimate the area of fractures generated from fracturing operations. We will conduct a series of modeling studies and present a methodology using typical transient pressure responses, simulated by the numerical model, to estimate fracture areas created or to quantity hydraulic fractures with traditional well testing technology. The type curves of pressure transients from this study can be used to quantify hydraulic fractures in field application.

  2. Safety analysis of RBMK design basis accidents by coupled neutronics thermal hydraulics codes QUABOX/CUBBOX-ATHLET and SADCO-ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Clemente, M.; Langenbuch, S.; Velkov, K. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Garching (Germany); Kuznetcov, P.; Panin, V.; Roghdestvensky, M.; Sakharova, T.; Stenbock, I. [Research and Development Inst. of Power Engineering (RDIPE), Moscow (Russian Federation)

    2006-07-01

    One of the most important issues of safety analysis is the validation and verification (V and V) of coupled neutronics/thermal-hydraulics codes used to study NPP transients and accidental processes. Risks and difficulties of a straight experimental investigation of such regimes of operating NPPs for V and V purposes give rise to the application of coupled 3D code packages cross-verification. - Codes for V and V: Thermal hydraulics and fluid mechanics calculations were performed by the ATHLET code (Analysis of Thermal-hydraulics of Leaks and Transient). Right and left loops of RBMK Main Coolant Circuit (MCC) connected by a steam line system were simulated. The following elements of each loop are reproduced as one group each: 2 Steam Drums, Downcomers and 3 Main Coolant Pumps. 22 Distribution Group Headers are modeled as one group for RIAs; however for LOCAs accidental and neighbouring channels are described separately. Fuel Channels (and the Lower Water Lines) are grouped according to the reactor core power distribution. In a first approach the core power in the ATHLET2.0A input model was simulated by a point kinetics model, which was replaced later by the 3D-core models SADCO from NIKIET and QUABOX/CUBBOX (Q/C) from GRS. - Results of Calculations: A comparison of Q/C and SADCO-ATHLET calculations for the NPP Kursk-1 are presented as example of 3D coupled codes cross-verification, showing a rod withdrawal accident. In the ATHLET model the RBMK core is represented by 6 regions, however the fuel channels around the accidental rod were analyzed separately, resulting in a total number of 79 thermal hydraulic channels in ATHLET to describe the core. The detailed description of modernized control and protection system (CPS) was included in the model. - Conclusions: 1. The results show that coupled codes cross-verification provides a reliable methodology of testing such codes for transient and accident analyses. 2. ATHLET is an effective calculational tool for analyses of

  3. 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)

  4. Joint Estimation of Hydraulic and Poroelastic Parameters from a Pumping Test.

    Science.gov (United States)

    Berg, Steven J; Illman, Walter A; Mok, Chin Man W

    2015-01-01

    The coupling of hydraulic and poroelastic processes is critical in predicting processes involving the deformation of the geologic medium in response to fluid extraction or injection. Numerical models that consider the coupling of hydraulic and poroelastic processes require the knowledge of relevant parameters for both aquifer and aquitard units. In this study, we jointly estimated hydraulic and poroelastic parameters from pumping test data exhibiting "reverse water level fluctuations," known as the Noordbergum effect, in aquitards adjacent to a pumped aquifer. The joint estimation was performed by coupling BIOT2, a finite element, two-dimensional, axisymmetric, groundwater model that considers poroelastic effects with the parameter estimation code PEST. We first tested our approach using a synthetic data set with known parameters. Results of the synthetic case showed that for a simple layered system, it was possible to reproduce accurately both the hydraulic and poroelastic properties for each layer. We next applied the approach to pumping test data collected at the North Campus Research Site (NCRS) on the University of Waterloo (UW) campus. Based on the detailed knowledge of stratigraphy, a five-layer system was modeled. Parameter estimation was performed by: (1) matching drawdown data individually from each observation port and (2) matching drawdown data from all ports at a single well simultaneously. The estimated hydraulic parameters were compared to those obtained by other means at the site yielding good agreement. However, the estimated shear modulus was higher than the static shear modulus, but was within the range of dynamic shear modulus reported in the literature, potentially suggesting a loading rate effect. © 2014, National GroundWater Association.

  5. Thermal Performance Testing of Cryogenic Insulation Systems

    Science.gov (United States)

    Fesmire, James E.; Augustynowicz, Stan D.; Scholtens, Brekke E.

    2007-01-01

    Efficient methods for characterizing thermal performance of materials under cryogenic and vacuum conditions have been developed. These methods provide thermal conductivity data on materials under actual-use conditions and are complementary to established methods. The actual-use environment of full temperature difference in combination with vacuum-pressure is essential for understanding insulation system performance. Test articles include solids, foams, powders, layered blankets, composite panels, and other materials. Test methodology and apparatus design for several insulation test cryostats are discussed. The measurement principle is liquid nitrogen boil-off calorimetry. Heat flux capability ranges from approximately 0.5 to 500 watts per square meter; corresponding apparent thermal conductivity values range from below 0.01 up to about 60 mW/m- K. Example data for different insulation materials are also presented. Upon further standardization work, these patented insulation test cryostats can be available to industry for a wide range of practical applications.

  6. Hydraulic Testing of Silurian and Ordovician Strata at the Bruce Site

    Science.gov (United States)

    Beauheim, R. L.; Avis, J. D.; Chace, D. A.; Roberts, R. M.; Toll, N. J.

    2009-05-01

    Ontario Power Generation is proposing a Deep Geologic Repository (DGR) for the long-term management of its Low and Intermediate Level Radioactive Waste (L&ILW) within a Paleozoic-age sedimentary sequence beneath the Bruce Site near Tiverton, Ontario, Canada. The concept envisions that the DGR would be excavated at a depth of approximately 680 m within the Ordovician Cobourg Formation, a massive, dense, argillaceous limestone. A key attribute of the Bruce site is the extremely low permeabilities associated with the thick Ordovician carbonate and argillaceous bedrock formations that will host and enclose the DGR. Such rock mass permeabilities are thought sufficiently low to contribute toward or govern a diffusion-dominated transport regime. To support this concept, hydraulic testing was performed in 2008 and 2009 in two deep boreholes at the proposed repository site, DGR-3 and DGR-4. The hydraulic testing was performed using a straddle-packer tool with a 30.74-m test interval. Sequential tests were performed over the entire open lengths of the boreholes from the F Unit of the Silurian Salina Formation into the Ordovician Gull River Formation, a distance of approximately 635 m. The tests consisted primarily of pressure-pulse tests, with a few slug tests performed in several of the higher permeability Silurian units. The tests are analyzed using the nSIGHTS code, which allows the entire pressure history a test interval has experienced since it was penetrated by the drill bit to be included in the test simulation. nSIGHTS also allows the model fit to the test data to be optimized over an n-dimensional parameter space to ensure that the final solution represents a true global minimum rather than simply a local minimum. The test results show that the Ordovician-age strata above the Coboconk Formation (70+ m below the Cobourg) have average horizontal hydraulic conductivities of 1E-13 m/s or less. Coboconk and Gull River hydraulic conductivities are as high as 1E-11 m

  7. Geomechanical, Hydraulic and Thermal Characteristics of Deep Oceanic Sandy Sediments Recovered during the Second Ulleung Basin Gas Hydrate Expedition

    Directory of Open Access Journals (Sweden)

    Yohan Cha

    2016-09-01

    Full Text Available This study investigates the geomechanical, hydraulic and thermal characteristics of natural sandy sediments collected during the Ulleung Basin gas hydrate expedition 2, East Sea, offshore Korea. The studied sediment formation is considered as a potential target reservoir for natural gas production. The sediments contained silt, clay and sand fractions of 21%, 1.3% and 77.7%, respectively, as well as diatomaceous minerals with internal pores. The peak friction angle and critical state (or residual state friction angle under drained conditions were ~26° and ~22°, respectively. There was minimal or no apparent cohesion intercept. Stress- and strain-dependent elastic moduli, such as tangential modulus and secant modulus, were identified. The sediment stiffness increased with increasing confining stress, but degraded with increasing strain regime. Variations in water permeability with water saturation were obtained by fitting experimental matric suction-water saturation data to the Maulem-van Genuchen model. A significant reduction in thermal conductivity (from ~1.4–1.6 to ~0.5–0.7 W·m−1·K−1 was observed when water saturation decreased from 100% to ~10%–20%. In addition, the electrical resistance increased quasi-linearly with decreasing water saturation. The geomechanical, hydraulic and thermal properties of the hydrate-free sediments reported herein can be used as the baseline when predicting properties and behavior of the sediments containing hydrates, and when the hydrates dissociate during gas production. The variations in thermal and hydraulic properties with changing water and gas saturation can be used to assess gas production rates from hydrate-bearing deposits. In addition, while depressurization of hydrate-bearing sediments inevitably causes deformation of sediments under drained conditions, the obtained strength and stiffness properties and stress-strain responses of the sedimentary formation under drained loading conditions

  8. Thermal-Transient Testing Of Turbine Blades

    Science.gov (United States)

    Wagner, William R.; Pidcoke, Louis H.

    1990-01-01

    Testing apparatus applies pulses of heat to turbine blade to determine resistance to thermal fatigue. Uses nonintrusive inductive heating and records distribution of temperature on blade with infrared video camera. Allows precise control of heating and cooling. Designed for testing blades used in advanced high-pressure, high-temperature turbines.

  9. Advanced Testing Method for Ground Thermal Conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xiaobing [ORNL; Clemenzi, Rick [Geothermal Design Center Inc.; Liu, Su [University of Tennessee (UT)

    2017-04-01

    A new method is developed that can quickly and more accurately determine the effective ground thermal conductivity (GTC) based on thermal response test (TRT) results. Ground thermal conductivity is an important parameter for sizing ground heat exchangers (GHEXs) used by geothermal heat pump systems. The conventional GTC test method usually requires a TRT for 48 hours with a very stable electric power supply throughout the entire test. In contrast, the new method reduces the required test time by 40%–60% or more, and it can determine GTC even with an unstable or intermittent power supply. Consequently, it can significantly reduce the cost of GTC testing and increase its use, which will enable optimal design of geothermal heat pump systems. Further, this new method provides more information about the thermal properties of the GHEX and the ground than previous techniques. It can verify the installation quality of GHEXs and has the potential, if developed, to characterize the heterogeneous thermal properties of the ground formation surrounding the GHEXs.

  10. RDS - A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

    Science.gov (United States)

    Salim, Mohd Faiz; Roslan, Ridha; Ibrahim, Mohd Rizal Mamat @

    2014-02-01

    Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBI-MOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges.

  11. Pumping-Test Evaluation of Fault-Zone Hydraulic Properties in a Fractured Sandstone

    Science.gov (United States)

    Johnson, N. M.

    2014-12-01

    Subzones of both reduced and enhanced permeability are often ascribed to fault zones, consistent with a fault-core/damage-zone conceptualization, with associated implications for assessing potential contaminant transport. Within this context, a 31-day pumping test was conducted in relation to a relatively minor, 2000 m long fault zone cutting fractured Cretaceous sandstone interbedded with siltstone and shale at a groundwater remediation site in the Simi Hills of southern California during March-April 2013. Our objective was to evaluate the potential hydrogeologic influence of the fault zone on groundwater movement across and along it by observing the spatial patterns of drawdown and estimated hydraulic properties. A 122 m deep open borehole was pumped at a constant rate of approximately 112 L/min while monitoring hydraulic heads in 14 observation wells, two completed with multi-level systems, within 750 m of the pumping well. Hydraulic heads were monitored for more than 9 months before, during, and after the test. Prior to the test, we used the site's three-dimensional, equivalent-porous-media groundwater flow model to anticipate the potential response of alternative fault-zone permeability structures. The results suggest that the fault zone may be slightly more permeable (by a factor of about 2 or less) and less confined than the fractured sandstone away from the fault, and is not a significant barrier to groundwater flow across it. Within the areal extent of observed drawdown, the site's hydrostratigraphic structures exhibited a relatively greater hydraulic influence. The pattern and magnitude of observed drawdown lie within the range of pre-test model simulations, and the test results are now being used to revise and recalibrate the model.

  12. IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bostelmann, F. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained). SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on

  13. Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook (and others)

    2007-08-15

    This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the

  14. Research on nanosatellite thermal cycling test applicability

    Directory of Open Access Journals (Sweden)

    Li Xiyuan

    2017-01-01

    Full Text Available In order to verify the spacecraft performance in extreme temperature and vacuum, and to screen spacecraft early defect, generally spacecraft TV (Thermal Vacuum test should be carried out before launch. Designed in small size and with low cost, nanosatellite is made from a large number of COTS (Commercial off the shelf components; therefore, the test should be low-cost, simple and quick. With the intention of screen out early defects of the product in lower cost, nanosatellite developers usually use TC (Thermal Cycling test to partially replace the TV test because TV test is more expensive. However, due to the air convection, TC test is different from TV test in heat transfer characteristics, which may be over-test or short-test in TC test. This paper aims to explore the applicability of different nanosatellites in TC/TV test. Using rule number analysis method, Heat Transfer model in vacuum and ambient environment has been built to analyse the characteristics of heat transfer under different temperature and characteristic length, and to deliver the recommended limits on using TC test instead of the TV test. The CFD and test methods are applied to verify the rule number analysis above.

  15. Hydraulic Testing of Salado Formation Evaporites at the Waste Isolation Pilot Plant Site: Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Beauheim, Richard L.; Domski, Paul S.; Roberts, Randall M.

    1999-07-01

    This report presents interpretations of hydraulic tests conducted in bedded evaporates of the Salado Formation from May 1992 through May 1995 at the Waste Isolation Pilot Plant (WIPP) site in southeastern New Mexico. The WIPP is a US Department of Energy research and development facility designed to demonstrate safe disposal of transuranic wastes from the nation's defense programs. The WIPP disposal horizon is located in the lower portion of the Permian Salado Formation. The hydraulic tests discussed in this report were performed in the WIPP underground facility by INTERA inc. (now Duke Engineering and Services, Inc.), Austin, Texas, following the Field Operations Plan and Addendum prepared by Saulnier (1988, 1991 ) under the technical direction of Sandia National Laboratories, Albuquerque, New Mexico.

  16. Neutronic and thermal-hydraulic calculations for the AP-1000 NPP with the MCNP6 and SERPENT codes

    Energy Technology Data Exchange (ETDEWEB)

    Stefani, Giovanni Laranjo; Maiorino, Jose R.; Santos, Thiago A., E-mail: giovanni.laranjo@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais; Rossi, Pedro R., E-mail: pedro.russorossi@gmail.com [FERMIUM - Tecnologia Nuclear, Sao Paulo, SP (Brazil)

    2015-07-01

    The AP-1000 is an evolutionary PWR reactor designed as an evolution of the AP-600 project. The reactor is already pre-licensed by NRC, and is considered to have achieved high standards of safety, possible short construction time and good economic competitiveness. The core is a 17x17 typical assembly using Zirlo as cladding, 3 different enrichment regions, and is controlled by boron, control banks, and burnable poison. The expected fuel final burnup is 62 MWD/ton U and a cycle of 18 months. In this paper we present results for neutronic and thermal-hydraulic calculations for the AP-1000. We use the MCNP6 and SERPENT codes to calculate the first cycle of operation. The calculated parameters are K{sub eff} at BOL and EOL and its variation with burnup and neutron flux, and reactivity coefficients. The production of transuranic elements such as Pu-239 and Pu-241, and burning fuel are calculated over time. In the work a complete reactor was burned for 450 days with no control elements, boron or burnable poison were considered, these results were compared with data provided by the Westinghouse. The results are compared with those reported in the literature. A simple thermal hydraulic analysis allows verification of thermal limits such as fuel and cladding temperatures, and MDNB. (author)

  17. Development of a 1D thermal-hydraulic analysis code for once-through steam generator in SMRs using straight tubes

    Energy Technology Data Exchange (ETDEWEB)

    Park, Youngjae; Kim, Iljin; Kim, Hyungdae [Kyung Hee University, Yongin (Korea, Republic of)

    2015-10-15

    Diverse integral/small-modular reactors (SMRs) have been developed. Once-through steam generator (OTSG) which generates superheated steam without steam separator and dryer was used in the SMRs to reduce volume of steam generator. It would be possible to design a new steam generator with best estimate thermal-hydraulic codes such as RELAP and MARS. However, it is not convenience to use the general purpose thermal-hydraulic analysis code to design a specific component of nuclear power plants. A widely used simulation tool for thermal-hydraulic analysis of drum-type steam generators is ATHOS, which allows 3D analysis. On the other hand, a simple 1D thermal-hydraulic analysis code might be accurate enough for the conceptual design of OTSG. In this study, thermal-hydraulic analysis code for conceptual design of OTSG was developed using 1D homogeneous equilibrium model (HEM). A benchmark calculation was also conducted to verify and validate the prediction accuracy of the developed code by comparing with the analysis results with MARS. Finally, conceptual design of OTSG was conducted by the developed code. A simple 1D thermal-hydraulic analysis code was developed for the purpose of conceptual design OTSG for SMRs. A set of benchmark calculations was conducted to verify and validate the analysis accuracy of the developed code by comparing results obtained with a best-estimated thermal-hydraulic analysis code, MARS. Finally, analysis of two different OTSG design concepts with superheating and recirculation was demonstrated using the developed code.

  18. Design and optimization of the WEST ICRH antenna front face components based on thermal and hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Zhaoxi, E-mail: chenzx@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); Vulliez, Karl [Laboratoire d’étanchéité, DEN/DTEC/SDTC, Commissariat à l’énergie atomique et aux énergies alternatives, 2 rue James Watt, 26700 Pierrelatte (France); Ferlay, Fabien; Martinez, André; Mollard, Patrick; Hillairet, Julien; Doceul, Louis; Bernard, Jean-Michel; Larroque, Sébastien; Helou, Walid [CEA, IRFM, F-13108, Saint-Paul-Lez-Durance (France); Song, Yuntao; Yang, Qingxi; Wang, Yongsheng [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2015-05-15

    Highlights: • Three ICRH antennas are designed to realize continuous-wave operation. • Fully active cooling structure is designed which takes the balance of structure safety and cooling performance. • High cooling efficiency is achieved for the current cooling circuit design based on the thermal-hydraulic simulation. - Abstract: The WEST (Tungsten (W) Environment in Steady-state Tokamak) is an upgrade of Tore-Supra (TS) which aims it into an X-point magnetic configuration tokamak equipped with an actively cooled tungsten divertor. To be a platform of ITER technologies of high heat flux components testing, three sets of Ion Cyclotron Resonant Heating (ICRH) antennas have been designed to inject 9 MW during 30 s or 3 MW during 1000 s. The antenna design is based on a load resilient prototype successfully tested in Tore Supra in 2007. In order to allow continuous-wave (CW) operations, the mechanical design of the WEST ICRH antenna is emphasized on its cooling performances by designing fully active cooling structure. Two kinds of cooling water loops are used, with temperature and pressure of 70 °C/30 bar and 25 °C/5.2 bar, respectively. The hot water loop is used for the Faraday screen (FS) and the housing box (HB), while the cold water loop is used for the straps, the matching capacitors and the impedance transformer. To enhance the heat removal ability and control the pressure drop, the cooling channels in the FS and HB are drilled directly and parallel connected as much as possible. By performing the hydraulic–thermal analysis, the lack of cooling efficiency was found in the front face of lateral collector where 1 MW/m{sup 2} is imposed and fluid dead zones were found in some of the bars. After optimization, the cooling performance of the cooling circuit increased significantly. With a mass flow rate of 2.5 kg/s, the total pressure drop is 3.1 bar, and the peak temperatures on the FS and HB are 500 °C and 261 °C, respectively. Besides, no cavitation is

  19. Final Report - Hydraulic Conductivity with Depth for Underground Test Area (UGTA) Wells

    Energy Technology Data Exchange (ETDEWEB)

    P. Oberlander; D. McGraw; C. Russell

    2007-10-31

    Hydraulic conductivity with depth has been calculated for Underground Test Area (UGTA) wells in volcanic tuff and carbonate rock. The following wells in volcanic tuff are evaluated: ER-EC-1, ER-EC-2a, ER-EC-4, ER-EC-5, ER-5-4#2, ER-EC-6, ER-EC-7, and ER-EC-8. The following wells in carbonate rock are evaluated: ER-7-1, ER-6-1, ER-6-1#2, and ER-12-3. There are a sufficient number of wells in volcanic tuff and carbonate rock to associate the conductivity values with the specific hydrogeologic characteristics such as the stratigraphic unit, hydrostratigraphic unit, hydrogeologic unit, lithologic modifier, and alteration modifier used to describe the hydrogeologic setting. Associating hydraulic conductivity with hydrogeologic characteristics allows an evaluation of the data range and the statistical distribution of values. These results are relevant to how these units are considered in conceptual models and represented in groundwater models. The wells in volcanic tuff illustrate a wide range of data values and data distributions when associated with specific hydrogeologic characteristics. Hydraulic conductivity data within a hydrogeologic characteristic can display normal distributions, lognormal distributions, semi-uniform distribution, or no identifiable distribution. There can be multiple types of distributions within a hydrogeologic characteristic such as a single stratigraphic unit. This finding has implications for assigning summary hydrogeologic characteristics to hydrostratigraphic and hydrogeologic units. The results presented herein are specific to the hydrogeologic characteristic and to the wells used to describe hydraulic conductivity. The wells in carbonate rock are associated with a fewer number of hydrogeologic characteristics. That is, UGTA wells constructed in carbonate rock have tended to be in similar hydrogeologic materials, and show a wide range in hydraulic conductivity values and data distributions. Associations of hydraulic conductivity and

  20. Parameter estimation of soil hydraulic and thermal property functions for unsaturated porous media using the HYDRUS-2D code

    Directory of Open Access Journals (Sweden)

    Nakhaei Mohammad

    2014-03-01

    Full Text Available Knowledge of soil hydraulic and thermal properties is essential for studies involving the combined effects of soil temperature and water input on water flow and redistribution processes under field conditions. The objective of this study was to estimate the parameters characterizing these properties from a transient water flow and heat transport field experiment. Real-time sensors built by the authors were used to monitor soil temperatures at depths of 40, 80, 120, and 160 cm during a 10-hour long ring infiltration experiment. Water temperatures and cumulative infiltration from a single infiltration ring were monitored simultaneously. The soil hydraulic parameters (the saturated water content θ s, empirical shape parameters α and n, and the saturated hydraulic conductivity Ks and soil thermal conductivity parameters (coefficients b1, b2, and b3 in the thermal conductivity function were estimated from cumulative infiltration and temperature measurements by inversely solving a two-dimensional water flow and heat transport using HYDRUS-2D. Three scenarios with a different, sequentially decreasing number of optimized parameters were considered. In scenario 1, seven parameters (θ s, Ks, α, n, b1, b2, and b3 were included in the inverse problem. The results indicated that this scenario does not provide a unique solution. In scenario 2, six parameters (Ks, α, n, b1, b2, and b3 were included in the inverse problem. The results showed that this scenario also results in a non-unique solution. Only scenario 3, in which five parameters (α, n, b1, b2, and b3 were included in the inverse problem, provided a unique solution. The simulated soil temperatures and cumulative infiltration during the ring infiltration experiment compared reasonably well with their corresponding observed values.

  1. Core Thermal-Hydraulic Design of a Sodium Cooled Fast Reactor for the U/TRU Fuel Modification

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Sun Rock; Cho, Chung Ho; Kim, Young Gyun; Song, Hoon; Park, Won Seok; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    The Korea Atomic energy Research Institute (KAERI) has been developing an advanced SFR design technology with the final goal of constructing a demonstration plant by 2028. The main objective of the SFR demonstration plant is to verify TRU metal fuel performance, large-scale reactor operation, and transmutation ability of high-level wastes. However, in the early stage, the SFR will run on low enriched uranium fuel due to a lack of TRU fuel qualification. After sequential evaluations of the fuel performance, the fissile fuel material will transform from uranium to LTRU (LWR-TRU), and then finally to MTRU (Mixed TRU of LTRU and recycled TRU). At the same time, the core configurations will be modified to meet the nuclear design requirements. Therefore, there is also a strong need to ensure a proper cooling capability during modifications of the entire core. In this work, the core thermal-hydraulic design for U/TRU fuel modification is performed using the SLTHEN (Steady-State LMR Thermal-Hydraulic Analysis Code Based on ENERGY Model) code. As the power distribution in a reactor core is not uniform, it requires a suitable flow allocation to each assembly. There are two ways of allocating the flow rates depending on the orifice positions. The inner officering scheme locates orifice plates in the lower part of the fuel assembly. Therefore, it is possible that the flow distribution is redesigned according to the core configurations. On the other hand, the outer officering scheme fixes orifice plates within the receptacle body throughout the entire plant lifetime. This has the advantage lower of fabrication costs and operating errors but included insufficient design flexibility. This paper provides comparative studies of orifice position for the core thermal-hydraulic design

  2. Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Giannetti, Fabio; Vitale Di Maio, Damiano; Naviglio, Antonio; Caruso, Gianfranco, E-mail: gianfranco.caruso@uniroma1.it

    2016-08-15

    Highlights: • LOOP thermal-hydraulic transient analysis for lead-cooled fast reactors. • Passive decay heat removal system concept to avoid lead freezing. • Solution developed for the diversification of the decay heat removal functions. • RELAP5 vs. RELAP5-3D comparison for lead applications. - Abstract: Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

  3. 3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Langenbuch, S.; Velkov, K. [GRS, Garching (Germany); Lizorkin, M. [Kurchatov-Institute, Moscow (Russian Federation)] [and others

    1997-07-01

    This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

  4. Laser thermal shock and fatigue testing system

    Science.gov (United States)

    Fantini, Vincenzo; Serri, Laura; Bianchi, P.

    1997-08-01

    Thermal fatigue consists in repeatedly cycling the temperature of a specimen under test without any other constraint and stopping the test when predefined damage aspects. The result is a lifetime in terms of number of cycles. The parameters of the thermal cycle are the following: minimum and maximum temperature, time of heating, of cooling and time at high or at low temperature. When the temperature jump is very big and fast, phenomena of thermal shock can be induced. Among the numerous techniques used to perform these tests, the laser thermal fatigue cycling is very effective when fast heating of small and localized zones is required. That's the case of test performed to compare new and repaired blades of turbogas machines or components of combustion chambers of energy power plants. In order to perform these tests a thermal fatigue system, based on 1 kW Nd-YAG laser as source of heating, has been developed. The diameter of the heated zone of the specimen irradiated by the laser is in the range 0.5 - 20 mm. The temperatures can be chosen between 200 degree(s)C and 1500 degree(s)C and the piece can be maintained at high and/or low temperature from 0 s to 300 s. Temperature are measured by two sensors: a pyrometer for the high range (550 - 1500 degree(s)C) and a contactless thermocouple for the low range (200 - 550 degree(s)C). Two different gases can be blown on the specimen in the irradiated spot or in sample backside to speed up cooling phase. A PC-based control unit with a specially developed software performs PID control of the temperature cycle by fast laser power modulation. A high resolution vision system of suitable magnification is connected to the control unit to detect surface damages on the specimen, allowing real time monitoring of the tested zone as well as recording and reviewing the images of the sample during the test. Preliminary thermal fatigue tests on flat specimens of INCONEL 738 and HAYNES 230 are presented. IN738 samples, laser cladded by

  5. Thermal Test on Target with Pressed Disks

    Energy Technology Data Exchange (ETDEWEB)

    Woloshun, Keith Albert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dale, Gregory E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Olivas, Eric Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Romero, Frank Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dalmas, Dale Allen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Chemerisov, Sergey [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Gromov, Roman [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lowden, Rick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-05-27

    A thorough test of the thermal performance of a target for Mo99 production using solid Mo100 target to produce the Mo99 via a gamma-n reaction has previously been conducted at Argonne National Laboratory (ANL). The results are reported in “Zero Degree Line Mo Target Thermal Test Results and Analysis,” LANL report Number LA-UR-15-23134 dated 3/27/15. This target was comprised of 25 disks 1 mm thick and 12 mm in diameter, separated by helium coolant gaps 0.5 mm wide. The test reported in the above referenced report was conducted with natural Mo disks all cut from commercial rod. The production plant will have Mo100 disks pressed and sintered using a process being developed at Oak Ridge National Laboratory (ORNL). The structural integrity of press-and-sinter disks is of some concern. The test reported herein included 4 disks made by the ORNL process and placed in the high heat, and therefore high thermal stress, region of the target. The electron beam energy was 23 MeV for these tests. Beam spot size was 3.5 mm horizontal and 3 mm vertical, FWHM. The thermal stress test of pressed-and-sintered disks resulted in no mechanical failures. The induced thermal stresses were below yield stress for natural Mo, indicating that up to that stress state no inherent deficiencies in the mechanical properties of the fabricated disks were evident.

  6. Thermal insulation testing method and apparatus

    Science.gov (United States)

    Fesmire, James E. (Inventor); Augustynowicz, Stanislaw D. (Inventor)

    2004-01-01

    A test apparatus and method of its use for evaluating various performance aspects of a test specimen is disclosed. A chamber within a housing contains a cold mass tank with a contact surface in contact with a first surface of a test specimen. The first surface of the test specimen is spaced from the second surface of the test specimen by a thickness. The second surface of the test specimen is maintained at a desired warm temperature. The first surface is maintained at a constant temperature by a liquid disposed within the cold mass tank. A boil-off flow rate of the gas is monitored and provided to a processor along with the temperature of the first and second surfaces of the test specimen. The processor calculates thermal insulation values of the test specimen including comparative values for heat flux and apparent thermal conductivity (k-value). The test specimen may be placed in any vacuum pressure level ranging from about 0.01 millitorr to 1,000,000 millitorr with different residual gases as desired. The test specimen may be placed under a mechanical load with the cold mass tank and another factors may be imposed upon the test specimen so as to simulate the actual use conditions.

  7. Functional linear models to test for differences in prairie wetland hydraulic gradients

    Science.gov (United States)

    Greenwood, Mark C.; Sojda, Richard S.; Preston, Todd M.; Swayne, David A.; Yang, Wanhong; Voinov, A.A.; Rizzoli, A.; Filatova, T.

    2010-01-01

    Functional data analysis provides a framework for analyzing multiple time series measured frequently in time, treating each series as a continuous function of time. Functional linear models are used to test for effects on hydraulic gradient functional responses collected from three types of land use in Northeastern Montana at fourteen locations. Penalized regression-splines are used to estimate the underlying continuous functions based on the discretely recorded (over time) gradient measurements. Permutation methods are used to assess the statistical significance of effects. A method for accommodating missing observations in each time series is described. Hydraulic gradients may be an initial and fundamental ecosystem process that responds to climate change. We suggest other potential uses of these methods for detecting evidence of climate change.

  8. Thermal and hydraulic characteristics of the JEN-1 Reactor; Caracteristicas hidraulicas y termicas del Reactor JEN-1

    Energy Technology Data Exchange (ETDEWEB)

    Otra Otra, F.; Leira Rey, G.

    1971-07-01

    In this report an analysis is made of the thermal and hydraulic performances of the JEN-1 reactor operating steadily at 3 Mw of thermal power. The analysis is made separately for the core, main heat exchanger and cooling tower. A portion of the report is devoted to predict the performances of these three main components when and if the reactor was going to operate at a power higher than the maximum 3 Mw attainable today. Finally an study is made of the unsteady operation of the reactor, focusing the attention towards the pumping characteristics and the temperatures obtained in the fuel elements. Reference is made to several digital calculation programmes that nave been developed for such purpose. (Author) 21 refs.

  9. Development of Mitsubishi high thermal performance grid 2 - overview of the development and Dnb test results

    Energy Technology Data Exchange (ETDEWEB)

    Hoshi, M.; Imaizumi, M.; Mori, M. [Mitsubishi Heavy Industries Ltd., Kobe (Japan); Hori, K. [Mitsubishi Heavy Industries Ltd., Takasago, Hyogo (Japan); Ikeda, K. [Nuclear Development Corp., Tokai, Ibaraki (Japan)

    2001-07-01

    Spacer grid plays fundamental role in thermal performance of PWR fuel assembly. Grid spacer with higher thermal performance gives greater DNB (Departure from Nucleate Boiling) margin for the core. Mitsubishi has developed a prototype Zircaloy grid with higher thermal performance. In this paper, process of the development and DNB test results of the grid is presented. To achieve a goal to design grid with higher DNB performance, CFD (Computational Fluid Dynamics) and Freon DNB test are employed in the development. It is also concerned that the grid should be hydraulically compatible to existing grid. CFD is used in examining mixing capability and pressure drop for early stage of the development. Freon DNB test is used for preliminary checking of DNB performance for several design of the grids. After the final design is fixed, DNB test has been carried out at a high pressure / high temperature water test loop to verify the DNB performance. Also, hydraulic test has been done in a water test loop. The test results show that the grid has higher DNB performance and lower pressure loss coefficient compared with existing grid. It is also concluded that a combination of CFD and Freon DNB testing is successful tool for designing and development of grid. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  11. Neutronics - thermal-hydraulics coupling: application to the helium-cooled fast reactor; Couplage neutronique - thermohydraulique: application au reacteur a neutrons rapides refroidi a l'helium

    Energy Technology Data Exchange (ETDEWEB)

    Vaiana, F.

    2009-11-15

    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)

  12. Thermal shock testing of lapped optical glass

    Science.gov (United States)

    Zhang, Yingrui; Wu, Yuansun; Liu, Han; Lambropoulos, John C.

    2007-09-01

    We have measured and modeled the thermal shock fracture of the commercially available BK-7 borosilicate crown optical glass as a function of surface finish prior to thermal shock testing. For surfaces lapped with alumina abrasives in the range 5 μm to 40 μm, the critical temperature drop for fracture in thin disk samples increases with diminishing abrasive size, and changes from 123.7+/-1.1 °C (for surfaces lapped with 40 μm abrasives) to 140.2+/-2.8 °C (for surfaces lapped with 5 μm abrasives.) We correlate the measured thermal shock (critical) temperature drop with the glass thermal and mechanical properties, including the fracture toughness, and the depth of surface cracks induced by the lapping process. We distinguish between "severe" and "mild" thermal shock conditions in terms of the applicable heat transfer coefficient and Biot number. We estimate that the depth of the strength controlling cracks on the edge of the disk samples was about 55-70 μm.

  13. Application of an In-Line Contamination Monitoring Unit to the AHT-64 Hydraulic Test Stand.

    Science.gov (United States)

    1981-06-04

    Specificat ion Ior ain In-1 ine Contaminat ion Monit (,r in I it ) J I roe-standing, two-wheel ed cart for the contaminit ion monitor; l)wov,.r, in, II...levels of vibration because it is a diesel-powered pomp utnit. Vibration tests of the AHT-64 indicate normal vibration amplitude to be 2 C ’it a critical...Commnd (NAVAIRSYSCOM) rejected the free-standing, two-wheeled cart concept proposed for the monitor and opted for an AHT-64 hydraulic test stand

  14. Ensemble Kalman filter versus ensemble smoother for assessing hydraulic conductivity via tracer test data assimilation

    Directory of Open Access Journals (Sweden)

    E. Crestani

    2013-04-01

    Full Text Available Estimating the spatial variability of hydraulic conductivity K in natural aquifers is important for predicting the transport of dissolved compounds. Especially in the nonreactive case, the plume evolution is mainly controlled by the heterogeneity of K. At the local scale, the spatial distribution of K can be inferred by combining the Lagrangian formulation of the transport with a Kalman-filter-based technique and assimilating a sequence of time-lapse concentration C measurements, which, for example, can be evaluated on site through the application of a geophysical method. The objective of this work is to compare the ensemble Kalman filter (EnKF and the ensemble smoother (ES capabilities to retrieve the hydraulic conductivity spatial distribution in a groundwater flow and transport modeling framework. The application refers to a two-dimensional synthetic aquifer in which a tracer test is simulated. Moreover, since Kalman-filter-based methods are optimal only if each of the involved variables fit to a Gaussian probability density function (pdf and since this condition may not be met by some of the flow and transport state variables, issues related to the non-Gaussianity of the variables are analyzed and different transformation of the pdfs are considered in order to evaluate their influence on the performance of the methods. The results show that the EnKF reproduces with good accuracy the hydraulic conductivity field, outperforming the ES regardless of the pdf of the concentrations.

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

    Energy Technology Data Exchange (ETDEWEB)

    Monti, Lanfranco, E-mail: lanfranco.monti@gmail.co [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtzplatz 1, 76344 Eggenstein-Leopoldshafen (Germany); Starflinger, Joerg, E-mail: joerg.starflinger@ike.uni-stuttgart.d [Universitaet Stuttgart, Institut fuer Kernenergetik und Energiesysteme, Pfaffenwaldring 31, 70569 Stuttgart (Germany); Schulenberg, Thomas, E-mail: schulenberg@kit.ed [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtzplatz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2011-05-15

    Highlights: Advanced analysis and design techniques for innovative reactors are addressed. Detailed investigation of a 3 pass core design with a multi-physics-scales tool. Coupled 40-group neutron transport/equivalent channels TH core analyses methods. Multi-scale capabilities: from equivalent channels to sub-channel pin-by-pin study. High fidelity approach: reduction of conservatism involved in core simulations. - Abstract: The High Performance Light Water Reactor (HPLWR) is a thermal spectrum nuclear reactor cooled and moderated with light water operated at supercritical pressure. It is an innovative reactor concept, which requires developing and applying advanced analysis tools as described in the paper. The relevant water density reduction associated with the heat-up, together with the multi-pass core design, results in a pronounced coupling between neutronic and thermal-hydraulic analyses, which takes into account the strong natural influence of the in-core distribution of power generation and water properties. The neutron flux gradients within the multi-pass core, together with the pronounced dependence of water properties on the temperature, require to consider a fine spatial resolution in which the individual fuel pins are resolved to provide precise evaluation of the clad temperature, currently considered as one of the crucial design criteria. These goals have been achieved considering an advanced analysis method based on the usage of existing codes which have been coupled with developed interfaces. Initially neutronic and thermal-hydraulic full core calculations have been iterated until a consistent solution is found to determine the steady state full power condition of the HPLWR core. Results of few group neutronic analyses might be less reliable in case of HPLWR 3-pass core than for conventional LWRs because of considerable changes of the neutron spectrum within the core, hence 40 groups transport theory has been preferred to the usual 2 groups

  16. Deep hydraulic tests in a large earth-slide rich in clay

    Science.gov (United States)

    Ronchetti, Francesco; Piccinini, Leonardo; Deiana, Manuela; Corsini, Alessandro

    2017-04-01

    Different hydraulic tests have been conducted and replicated in a large earth slide characterized by a landslide body that is rich in clay, has a mean thickness of 30 meters, and is located in the Northern Apennines, Italy. All the tests were performed to estimate the hydrogeological properties of the landslide and to design future mitigation measures. To define the geometry of the sliding mass, the stratigraphy in more than 15 boreholes was analyzed. The boreholes were subsequently equipped with inclinometers and open standpipe piezometers. According to the stratigraphy, the landslide body is characterized by the presence of gravel layers in a clay-rich matrix. This study compares the results from the different techniques applied to 2 boreholes, 5 open standpipe piezometers and 1 well. The number of tests performed for each test type were 31 slug tests (ST), 4 falling head tests (FT), 5 low-flow pumping tests (PT), 1 point dilution (PD) test, and 2 aquifer tests (AT). Moreover, the test data was evaluated with different solutions. The ST data was evaluated with the Hvorslev and KGS solutions; the FT data was evaluated with the AGI and Hvorslev solutions; the PT data was evaluated with the Muskat solutions; the AT data was evaluated with the Theis, Cooper-Jacob, Neuman, Moench and Tartakosky-Neuman solutions; and the PD test data was evaluated with the classical solution where Darcy velocity is calculated as a function of the rate of dilution. The results show that hydraulic conductivity (K), storage (S) and specific storage (Ss) vary in the horizontal plane and with the depth (K ranges between 1.0E-5 and 1.0E-8 m/s; S ranges between 4.0E-3 and 5E-5; and Ss ranges between 1.0E-3 and 3.0E-3 1/m). The horizontal and vertical variability is correlated with the lithologic heterogeneity highlighted by the borehole stratigraphy. Moreover, all the hydraulic tests conducted on the landslide body give highly consistent results. Comparison of results derived from different

  17. Simulation and Design of High-Speed Hydraulic Velocity Generator in Shock Test Machine

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Hyeong; Shul, Chang Won; Kim, Yoon Jae; Yang, Myung Seog [Agency for Defense Development, Daejeon (Korea, Republic of); Lee, Gyu Sub [RMS Technology Corp., Seoul (Korea, Republic of)

    2014-06-15

    Mechanical and electrical devices in various forms are used in many different fields. These can be exposed to external environmental factors such as shock. Therefore, a shock test machine is commonly used to test these devices and evaluate their shock resistance. In this test, the break-down or permanent deformation and malfunction of inner parts due to a high stress or acceleration can be evaluated. As part of a shock test machine, a velocity generator is needed to create shocks between objects. In this study, a hydraulic velocity generator was conceptually designed and an AMESim model was developed to simulate the velocity under different conditions. Simulation results using this model were compared with the test results from a reduced-size velocity generator, and we designed a velocity generator that fits the target payload and velocity using the simulation results.

  18. Test facility requirements for the thermal vacuum thermal balance test of the Cosmic Background Explorer Observatory

    Science.gov (United States)

    Milam, Laura J.

    1991-01-01

    The Cosmic Background Explorer Observatory (COBE) underwant a thermal vacuum thermal balance test in the Space Environment Simulator (SES). This was the largest and most complex test ever conducted at this facility. The 4 x 4 m (13 x 13 ft) spacecraft weighed approx. 2223 kg (4900 lbs) for the test. The test set up included simulator panels for the inboard solar array panels, simulator panels for the flight cowlings, Sun and Earth Sensor stimuli, Thermal Radio Frequency Shield heater stimuli and a cryopanel for thermal control in the Attitude Control System Shunt Dissipator area. The fixturing also included a unique 4.3 m (14 ft) diameter Gaseous Helium Cryopanel which provided a 20 K environment for the calibration of one of the spacecraft's instruments, the Differential Microwave Radiometer. This cryogenic panel caused extra contamination concerns and a special method was developed and written into the test procedure to prevent the high buildup of condensibles on the panel which could have led to backstreaming of the thermal vacuum chamber. The test was completed with a high quality simulated space environment provided to the spacecraft. The test requirements, test set up, and special fixturing are described.

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

    Energy Technology Data Exchange (ETDEWEB)

    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. Steady-state thermal hydraulic analysis and flow channel blockage accident analysis of JRR-4 silicide LEU core

    Energy Technology Data Exchange (ETDEWEB)

    Kaminaga, Masanori; Yamamoto, Kazuyoshi; Watanabe, Shukichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-09-01

    JRR-4 is a light water moderated and cooled, graphite reflected pool type research reactor using high enriched uranium (HEU) plate-type fuels. Its thermal power is 3.5 MW. The core conversion program from HEU fuel to uranium-silicon-aluminum (U{sub 3}Si{sub 2}-Al) dispersion type fuel (Silicide fuel) with low enriched uranium (LEU) is currently conducted at the JRR-4. This report describes about the steady-state thermal hydraulic analysis results and the flow channel blockage accident analysis result. In JRR-4, there are two operation mode. One is high power operation mode up to 3.5 MW, under forced convection cooling using the primary and the secondary cooling systems. The other is low power operation mode up to 200 kW, under natural circulation cooling between the reactor core and the reactor pool without the primary and the secondary cooling systems. For the analysis of the high power operation mode under forced convection cooling and the flow channel blockage accident, COOLOD code was used. On the other hand, for the analysis of low power operation under natural convection cooling, COOLOD-N2 code was used. From steady-state thermal hydraulic analysis results of both forced and natural convection cooling, fuel temperature, minimum DNBR etc. meet the design criteria and JRR-4 LEU silicide core has enough safety margin under normal operation conditions. Furthermore, flow channel blockage accident analysis results show that one channel flow blockage accident meet the safety criteria for accident conditions which have been established for JRR-4 LEU silicide core. (author)

  1. Understanding the Geometry of Connected Fracture Flow with Multiperiod Oscillatory Hydraulic Tests.

    Science.gov (United States)

    Sayler, Claire; Cardiff, Michael; Fort, Michael D

    2018-03-01

    An understanding of the spatial and hydraulic properties of fast preferential flow pathways in the subsurface is necessary in applications ranging from contaminant fate and transport modeling to design of energy extraction systems. One method for the characterization of fracture properties over interwellbore scales is Multiperiod Oscillatory Hydraulic (MOH) testing, in which the aquifer response to oscillatory pressure stimulations is observed. MOH tests were conducted on isolated intervals of wells in siliciclastic and carbonate aquifers in southern Wisconsin. The goal was to characterize the spatial properties of discrete fractures over interwellbore scales. MOH tests were conducted on two discrete fractured intervals intersecting two boreholes at one field site, and a nest of three piezometers at another field site. Fracture diffusivity estimates were obtained using analytical solutions that relate diffusivity to observed phase lag and amplitude decay. In addition, MOH tests were used to investigate the spatial extent of flow using different conceptual models of fracture geometry. Results indicated that fracture geometry at both field sites can be approximated by permeable two-dimensional fracture planes, oriented near-horizontally at one site, and near-vertically at the other. The technique used on MOH field data to characterize fracture geometry shows promise in revealing fracture network characteristics important to groundwater flow and transport. © 2017, National Ground Water Association.

  2. Integrated Cryogenic Propulsion Test Article Thermal Vacuum Hotfire Testing

    Science.gov (United States)

    Morehead, Robert L.; Melcher, J. C.; Atwell, Matthew J.; Hurlbert, Eric A.

    2017-01-01

    In support of a facility characterization test, the Integrated Cryogenic Propulsion Test Article (ICPTA) was hotfire tested at a variety of simulated altitude and thermal conditions in the NASA Glenn Research Center Plum Brook Station In-Space Propulsion Thermal Vacuum Chamber (formerly B2). The ICPTA utilizes liquid oxygen and liquid methane propellants for its main engine and four reaction control engines, and uses a cold helium system for tank pressurization. The hotfire test series included high altitude, high vacuum, ambient temperature, and deep cryogenic environments, and several hundred sensors on the vehicle collected a range of system level data useful to characterize the operation of an integrated LOX/Methane spacecraft in the space environment - a unique data set for this propellant combination.

  3. Oscillatory hydraulic testing as a strategy for NAPL source zone monitoring: Laboratory experiments

    Science.gov (United States)

    Zhou, YaoQuan; Cardiff, Michael

    2017-05-01

    Non-aqueous phase liquids (NAPLs) have a complex mode of transport in heterogeneous aquifers, which can result in pools and lenses of NAPLs (the ;source zone;) that are difficult to detect and can cause long-term contamination via slow dissolution into groundwater (the ;dissolved plume;). Characterizing the extent and evolution of NAPL contamination within the source zone is a useful strategy for designing and adapting appropriate remedial actions at many contaminated sites. As a NAPL flows into a given aquifer volume, the effective hydraulic conductivity (K) and specific storage (Ss) of the volume changes associated with the viscosity and compressibility of the impinging fluid, meaning that NAPL movement may be detectable with hydraulic testing. Recently, the use of oscillatory pumping tests - in which sinusoidal pumping variations are implemented and oscillatory pressure changes are detected at monitoring locations - has been suggested as a low-impact hydraulic testing strategy for characterizing aquifer properties (Cardiff et al., 2013; Zhou et al., 2016). Here, we investigate this strategy in an experimental laboratory sandbox where dyed vegetable oil is injected and allowed to migrate as a NAPL. Initial qualitative analyses demonstrate that measurable changes in pressure signal amplitude and phase provide clear evidence for NAPL plume emplacement and migration. Using the approach developed in Zhou et al. (2016), we then apply tomographic analyses to estimate the location of effective K changes (representing fluid changes) and their movement throughout time. This approach provides a method for monitoring ongoing NAPL movement without net extraction or injection of fluid, making it advantageous in field remediation applications.

  4. Interpretation and evaluation of monitoring results for main pipelines hydraulic pressure testing under non-isothermal conditions

    Science.gov (United States)

    Kraus, Ju. A.; Ivanov, R. N.; Grinevich, V. A.; Pakhotin, A. N.

    2017-08-01

    The paper examines hydraulic pressure testing of main oil pipelines. The research objective is to identify and describe the characteristic features of pressure changes in pressure testing of the main pipeline, caused by temperature changes. The notions on the interpretation and evaluation of the results for pipeline pressure testing under non-isothermal conditions are given, the number of pressure sampling points and temperature information being limited; and considering the elevation difference along the length of the test section (where applicable). A formula for calculating the fresh water volumetric expansion coefficient correlation with temperature and pressure is proposed. A method for interpreting the hydraulic pressure testing results is developed, considering the parameters spread effect on the coefficients of the volumetric expansion and the modulus of elasticity for the test fluid, as well as the coefficients of linear expansion and Young's modulus for pipe steel. The application of the method allows to monitor the hydraulic pressure testing for main oil pipelines.

  5. Conceptual Thermal Hydraulic Design of a 20MW Multipurpose Research Reactor (KAERI/VAEC joint study on a new research reactor for Vietnam)

    Energy Technology Data Exchange (ETDEWEB)

    Chae, Hee Taek; Seo, Chul Gyo; Park, Jong Hark; Park, Cheol [Kaeri, Daejeon (Korea, Republic of); Vinh, Le Vinh; Nghiem, Huynh Ton; Dang, Vo Doan Hai [Dalat Nuclear Research Reactor, Hanoi (Viet Nam)

    2007-08-15

    The conceptual thermal hydraulics design analyses for the 20 MW reference AHR core have been jointly performed by the KAERI and DNRI(VAEC). The preliminary core thermal hydraulic characteristics and safety margins for the AHR core were studied for various core flow rates, fuel assembly powers and core inlet temperatures. Statistical method was applied to the thermal hydraulic design of the reactor core. The MATRA{sub h} subchannel code has been applied to evaluate the thermal hydraulic performances of the AHR and the resulting thermal margins of the core under the forced convection cooling mode during a nominal power operation and the natural circulation mode during a reactor shutdown condition. In addition, typical accident analyses were carried out for a loss of flow accident by a primary pump seizure and a reactivity induced accident by a CAR rod withdrawal during a normal full power operation. The normal full power operation of the AHR was ensured with a sufficient safety margin for the onset of nucleate boiling phenomena. The AHR also had a sufficient natural circulation cooling capability to cool the core without the onset of nucleate boiling in the channel after a normal reactor shutdown and the anticipated transients. It was confirmed by the typical accident analyses that the AHR core was sufficiently protected from the loss of flow by the primary cooling pump seizure and the overpower transients by the CAR withdrawal from the MCHFR and fuel temperature points of view.

  6. Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, Hideki, E-mail: kamide.hideki@jaea.go.jp; Ohshima, Hiroyuki, E-mail: ohshima.hiroyuki@jaea.go.jp; Sakai, Takaaki, E-mail: sakai.takaaki@jaea.go.jp; Tanaka, Masaaki, E-mail: tanaka.masaaki@jaea.go.jp

    2017-02-15

    Highlights: • Thermal hydraulic issues for safety design criteria of sodium cooled fast reactors. • Measurement of velocity data in a subchannel surrounded by wire wrapped fuel-pins. • Statistical evaluation of core hot spot temperature during natural circulation. • Simulation of dynamics of molten fuel pool in a core disruptive accident. • V&V procedure of a multi-dimensional thermal hydraulic code on thermal striping. - Abstract: In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related R&D results on innovative technologies and lessons learned from Fukushima Dai-ichi nuclear power plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V&V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.

  7. Experimental Development and Demonstration of Ultrasonic Measurement Diagnostics for Sodium Fast Reactor Thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Tokuhiro, Akira; Jones, Byron

    2013-09-13

    This research project will address some of the principal technology issues related to sodium-cooled fast reactors (SFR), primarily the development and demonstration of ultrasonic measurement diagnostics linked to effective thermal convective sensing under normatl and off-normal conditions. Sodium is well-suited as a heat transfer medium for the SFR. However, because it is chemically reactive and optically opaque, it presents engineering accessibility constraints relative to operations and maintenance (O&M) and in-service inspection (ISI) technologies that are currently used for light water reactors. Thus, there are limited sensing options for conducting thermohydraulic measurements under normal conditions and off-normal events (maintenance, unanticipated events). Acoustic methods, primarily ultrasonics, are a key measurement technology with applications in non-destructive testing, component imaging, thermometry, and velocimetry. THis project would have yielded a better quantitative and qualitative understanding of the thermohydraulic condition of solium under varied flow conditions. THe scope of work will evaluate and demonstrate ultrasonic technologies and define instrumentation options for the SFR.

  8. Model studies of the vertical steam generator thermal-hydraulic characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Desyatun, V.F.; Moskvichev, V.F.; Ulasov, V.M.; Morozov, V.G.; Burkov, V.K.; Grebennikov, V.N.

    1984-03-01

    Results of investigations conducted to clarify the calculation technique and to test the workability of the main elements and units of the PGV-250 vertical steam generator of saturated steam are considered. The steam generating capacity of the plant is 1486 t/h, thermal power is 792 MW. Steam generation follows a multiple circulation scheme. The heat surface comprises 330-shields. The investigations are carried out with a model which reproduces all the main elements of the steam generator excluding the economizer section. The flow rates of feed water, generated steam and coolant of the first circuit as well as temperature, pressure and humidity of the generated steam past the separator are determined. The average heat transfer factors of the heat surface are calculated on the base of the data obtained and a conclusion is drawn on the correctness of the thermohydraulic calculation technique used in development of the PGV-250 steam generator design. Temperature pulsations and heat surface steaming are not observed. The steam humidity at the outlet and steam capture into sinking tubes are within permissible values.

  9. A mini slug test method for determination of a local hydraulic conductivity of an unconfined sandy aquifer

    DEFF Research Database (Denmark)

    Hinsby, Klaus; Bjerg, Poul Løgstrup; Andersen, Lars J.

    1992-01-01

    A new and efficient mini slug test method for the determination of local hydraulic conductivities in unconfined sandy aquifers is developed. The slug test is performed in a small-diameter (1 inch) driven well with a 0.25 m screen just above the drive point. The screened drive point can be driven...... from level to level and thereby establish vertical profiles of the hydraulic conductivity. The head data from the test well are recorded with a 10 mm pressure transducer, and the initial head difference required is established by a small vacuum pump. The method described has provided 274 spatially...... distributed measurements of a local hydraulic conductivity at a tracer test site at Vejen, Denmark. The mini slug test results calculated by a modified Dax slug test analysing method, applying the elastic storativity in the Dax equations instead of the specific yield, are in good accordance with the results...

  10. System for Testing Thermal Insulation of Pipes

    Science.gov (United States)

    Fesmire, James E.; Augustynowicz, Stanislaw D.; Nagy, Zoltan F.

    2003-01-01

    An apparatus and method have been developed for measuring the rates of leakage of heat into pipes carrying liquids, the purpose of the measurements being to quantify the thermal performance of the insulation system. The apparatus is designed primarily for testing pipes used to carry cryogenic liquids, but can also be used for measuring the thermal performance of other insulated pipes or piping systems. The basic measurement principle is straightforward: The outer surface of the pipe insulation is maintained at a fixed warmer temperature. The interior of the pipe is maintained in a narrow fixed lower-temperature range by means of a regular liquid (e.g., water) that is pumped through the pipe at a known flow rate or a cryogenic liquid (e.g., nitrogen) that is saturated at atmospheric pressure and replenished until steady-state conditions are achieved. In the case of water or another liquid pumped through, the inlet and outlet temperatures are measured and heat-leak power is calculated as the mass flow rate of the liquid multiplied by the specific heat of the liquid multiplied by the inlet-to-outlet temperature rise of the liquid. In the case of liquid nitrogen or another low-temperature boiling liquid, the heat-leak power is calculated as the rate of boil-off multiplied by the latent heat of vaporization of the liquid. Then the thermal-insulation performance of the pipe system can be calculated as a function of the measured heat-leak power, the inner and outer boundary temperatures, and the dimensions of the pipe. The apparatus can test as many as three pipes simultaneously. The pipes can have inner diameters up to .15 cm and outer diameters up to .20 cm. The lengths of the pipes may vary; typical lengths are of the order of 18 m. Two thermal guard boxes . one for each end of the pipe(s) under test . are used to make the inlet and outlet fluid connections to the pipe(s) (see figure). The connections include bellows that accommodate thermal expansion and contraction

  11. Uncertainties in determination of the hydraulic conductivity by physical model test

    Science.gov (United States)

    Barta, Eszter; Hajnal, Géza; Vasvári, Vilmos

    2013-04-01

    To determine the Darcy's coefficient of permeability are several methods available. Empirical and deterministic calculation methods were developed of which applicability and accuracy depend on the available data and the type of investigated soil. Both field and laboratory investigations are common. In practice of civil engineering it is most essential task prior excavation to determine this soil physical parameter for planning of dewatering systems. Field investigations play central role also in the determination of recoverable water resources. In practice it is not common that all data required for the field investigation - usually pumping test - and its evaluation are available, the well design and the conditions of the measurement do not meet those assumed in the theory. Due to information of poor quality and anomalous conditions the calculated coefficient of permeability and the seepage hydraulic parameters can differ from the real values. The aims of the investigations were to conduct laboratory model tests in different soil types, also in their layered structure and by different design of the pumping well, to evaluate their results supported by numerical modelling and to come to conclusions which can be helpful in the areas mentioned above. In the course of the measurements size fraction and features of the pumping well were varied in order to achieve realistic field conditions. A laboratory model integrated also the field experiences was created. A cylindrically symmetrical model with a ground plain of a quadrant, a radius of 1.325 m and a height of 1.0 m was used. Moreover by means of the investigation's results recommendations can be made for the layout of field tests (number of observation wells, distance of wells), for the type of the hydraulic test (conventional pumping test, single well test, slug test) and for the best applicable evaluation method.

  12. Physical and Hydraulic Properties of Rock Specimens from Grimsel Test Site, Switzerland

    Science.gov (United States)

    Zhang, M.; Takeda, M.

    2007-12-01

    The Grimsel Test Site (GTS) is located at an altitude of 1730 meters in the granite rock of Aar Massif in central Switzerland. It lies at a depth of around 450 meters beneath the surface and was established over 20 years ago by the National Cooperative for the Disposal of Radioactive Waste (Nagra) as a center for underground Research and Development (R&D) supporting a wide range of related research projects. Among many of them, an international cooperative project defined as the Long Term Diffusion (LTD) has been performing to obtain quantitative information on matrix diffusion in rock strata under in-situ conditions. A set of laboratory experiments is also planed to determine the physical, hydraulic and diffusive transport properties of rock specimens taken from the same test site. In this poster, we present the preliminary results of both physical and hydraulic properties of the rock specimens being tested, including the bulk density, porosity, specific surface area and pore distribution, microstructure, P and S wave velocities, electrical resistivity, air and water permeabilities. The results obtained in this study indicate that: 1) The porosity and wave velocities of Grimsel granite are relatively low compared to the average values of igneous rocks indicating that micro-cracks can potentially exist within grain minerals. 2) The air and water permeabilities obtained from the air and water permeability tests are consistent that illustrates the accuracy of both experiments. 3) Permeability is not a simple function of effective confining pressure. It is very sensitive to confining pressure especially at low pressure levels. Besides, the permeability is hysteretic depending on confining pressure, pore pressure and stress history. 4) Similar to the hydraulic properties, diffusive transport properties of rock specimens can also be significantly affected by the confining pressure conditions. Laboratory diffusion tests considering the effects of stress conditions are

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

  14. LMFBR in-core thermal-hydraulics: the state of the art and US research and development needs

    Energy Technology Data Exchange (ETDEWEB)

    Khan, E.U.

    1980-04-01

    A detailed critical review is presented of the literature relevant to predicting coolant flow and temperature fields in LMFBR core assemblies for nominal and non-nominal rod bundle geometries and reactor operating conditions. The review covers existing thermal-hydraulic models, computational methods, and experimental data useful for the design of an LMFBR core. The literature search made for this review included publications listed by Nuclear Science Abstracts and Energy Data Base as well as papers presented at key nuclear conferences. Based on this extensive review, the report discusses the accuracy with which the models predict flow and temperature fields in rod assemblies, identifying areas where analytical, experimental, and model development needs exist.

  15. Transient cases analyses of the TRIGA IPR-R1 using thermal hydraulic and neutron kinetic coupled codes

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

    Simulations and analyses of nuclear reactors have been improved by utilization of coupled thermal-hydraulic (TH) and neutron kinetics (NK) system codes especially to simulate transients that involve strong feedback effects between NK and TH. The TH-NK coupling technique was initially developed and used to simulate the behavior of power reactors; however, several coupling methodologies are now being applied for research reactors. This work presents the coupling methodology application between RELAP5 and PARCS codes using as a model the TRIGA IPR-R1 research reactor. Analyses of steady state and transient conditions and comparisons with results from simulations using only the RELAP5 code are being presented in this paper. (author)

  16. Thermal-hydraulic processes involved in loss of residual heat removal during reduced inventory operation. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Fletcher, C.D.; McHugh, P.R.; Naff, S.A.; Johnsen, G.W.

    1991-02-01

    This paper identifies the topics needed to understand pressurized water reactor response to an extended loss of residual heat removal event during refueling and maintenance outages. By identifying the possible plant conditions and cooling methods that would be used for each cooling mode, the controlling thermal-hydraulic processes and phenomena were identified. Controlling processes and phenomena include: gravity drain, core water boil-off, and reflux cooling processes. Important subcategories of the reflux cooling processes include: the initiation of reflux cooling from various plant conditions, the effects of air on reflux cooling, core level depression effects, issues regarding the steam generator secondaries, and the special case of boiler-condenser cooling with once-through steam generators. 25 refs., 6 figs., 1 tab.

  17. RELAP5-3D thermal hydraulic analysis of the target cooling system in the SPES experimental facility

    Science.gov (United States)

    Giardina, M.; Castiglia, F.; Buffa, P.; Palermo, G.; Prete, G.

    2014-11-01

    The SPES (Selective Production of Exotic Species) experimental facility, under construction at the Italian National Institute of Nuclear Physics (INFN) Laboratories of Legnaro, Italy, is a second generation Isotope Separation On Line (ISOL) plant for advanced nuclear physic studies. The UCx target-ion source system works at temperature of about 2273 K, producing a high level of radiation (105 Sv/h), for this reason a careful risk analysis for the target chamber is among the major safety issues. In this paper, the obtained results of thermofluid-dynamics simulations of accidental transients in the SPES target cooling system are reported. The analysis, performed by using the RELAP5-3D 2.4.2 qualified thermal-hydraulic system code, proves good safety performance of this system during different accidental conditions.

  18. Thermal Protection Test Bed Pathfinder Development Project

    Science.gov (United States)

    Snapp, Cooper

    2015-01-01

    In order to increase thermal protection capabilities for future reentry vehicles, a method to obtain relevant test data is required. Although arc jet testing can be used to obtain some data on materials, the best method to obtain these data is to actually expose them to an atmospheric reentry. The overprediction of the Orion EFT-1 flight data is an example of how the ground test to flight traceability is not fully understood. The RED-Data small reentry capsule developed by Terminal Velocity Aerospace is critical to understanding this traceability. In order to begin to utilize this technology, ES3 needs to be ready to build and integrate heat shields onto the RED-Data vehicle. Using a heritage Shuttle tile material for the heat shield will both allow valuable insight into the environment that the RED-Data vehicle can provide and give ES3 the knowledge and capability to build and integrate future heat shields for this vehicle.

  19. Molybdenum-99-producing 37-element fuel bundle neutronically and thermal-hydraulically equivalent to a standard CANDU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Nichita, E., E-mail: Eleodor.Nichita@uoit.ca; Haroon, J., E-mail: Jawad.Haroon@uoit.ca

    2016-10-15

    Highlights: • A 37-element fuel bundle modified for {sup 99}Mo production in CANDU reactors is presented. • The modified bundle is neutronically and thermal-hydraulically equivalent to the standard bundle. • The modified bundle satisfies all safety criteria satisfied by the standard bundle. - Abstract: {sup 99m}Tc, the most commonly used radioisotope in diagnostic nuclear medicine, results from the radioactive decay of {sup 99}Mo which is currently being produced at various research reactors around the globe. In this study, the potential use of CANDU power reactors for the production of {sup 99}Mo is investigated. A modified 37-element fuel bundle, suitable for the production of {sup 99}Mo in existing CANDU-type reactors is proposed. The new bundle is specifically designed to be neutronically and thermal-hydraulically equivalent to the standard 37-element CANDU fuel bundle in normal, steady-state operation and, at the same time, be able to produce significant quantities of {sup 99}Mo when irradiated in a CANDU reactor. The proposed bundle design uses fuel pins consisting of a depleted-uranium centre surrounded by a thin layer of low-enriched uranium. The new molybdenum-producing bundle is analyzed using the lattice transport code DRAGON and the diffusion code DONJON. The proposed design is shown to produce 4081 six-day Curies of {sup 99}Mo activity per bundle when irradiated in the peak-power channel of a CANDU core, while maintaining the necessary reactivity and power rating limits. The calculated {sup 99}Mo yield corresponds to approximately one third of the world weekly demand. A production rate of ∼3 bundles per week can meet the global demand of {sup 99}Mo.

  20. Thermal-hydraulic analysis of an irregular sector of the ITER vacuum vessel by means of CFD tools

    Energy Technology Data Exchange (ETDEWEB)

    Fradera, J., E-mail: jorge.fradera@idom.com [Idom Nuclear Services, Gran Vía Carlos III, 97 Bajos, 08028 Barcelona (Spain); Colomer, C.; Fabbri, M.; Martín, M.; Martínez-Sabán, E.; Zamora, I.; Alemán, A. [Idom Nuclear Services, Gran Vía Carlos III, 97 Bajos, 08028 Barcelona (Spain); Izquierdo, J. [F4E, Barcelona (Spain); Le Barbier, R.; Utin, Y. [ITER IO, Cadarache (France)

    2015-03-15

    Highlights: • 3D Geometry healing and simplification for CFD simulations. • Meshing of large domains for CFD simulations. • Meshing procedure for fluid–solid interface coupling. • Hydraulics of the ITER VV Irregular Sector number 2 (IrS#2). • Thermal-hydraulics of the ITER VV IrS#2. - Abstract: The present work exposes the 3D thermal-hydraulic analysis of the Irregular Sector number 2 (IrS#2) of the ITER Vacuum Vessel (VV) by means of CFD (computational fluid dynamics). IrS#2 geometry has been simplified and healed in order to be suitable for CFD analysis. A polyhedral cell based mesh has been generated so as to enhance accuracy and calculation stability. Nuclear heat deposition has been implemented through several subroutines and an in-house MCNP data converter. Water coolant and stainless steel shell are solved coupled as a steady-state conjugate heat transfer problem in order to assess the impact of the nuclear heat deposition on the IrS#2 cooling scheme. Hence, the IrS#2 is simulated as a whole without splitting the domain. Results show the total IrS#2 pressure drop as well as the flow and temperature distribution all over the IrS#2. Moreover, heat transfer coefficient has been calculated at the water–shell interface in order to assess the behavior of shell cooling scheme. Velocity magnitude in the water coolant has an average value of 2 cm/s and the inboard to outboard mass flow rate distribution is 10.2% and 89.8% respectively. Pressure drop, mainly at inlet and outlet ducts, is of 60.21 kPa. Temperature at the liquid–solid interface has an average value of 106.4 °C and the heat transfer coefficient (HTC) stays always above 638 W/(m{sup 2} °C), way above the limit of 500 W/(m{sup 2} °C). Shell temperature stays at an average value of 130.0 °C. Exposed results, with a significant importance regarding design and safety, give a valuable insight on current cooling scheme and system behavior for the IrS#2 of the ITER VV.

  1. Transient thermal, hydraulic, and mechanical analysis of a counter flow offset strip fin intermediate heat exchanger using an effective porous media approach

    Science.gov (United States)

    Urquiza, Eugenio

    This work presents a comprehensive thermal hydraulic analysis of a compact heat exchanger using offset strip fins. The thermal hydraulics analysis in this work is followed by a finite element analysis (FEA) to predict the mechanical stresses experienced by an intermediate heat exchanger (IHX) during steady-state operation and selected flow transients. In particular, the scenario analyzed involves a gas-to-liquid IHX operating between high pressure helium and liquid or molten salt. In order to estimate the stresses in compact heat exchangers a comprehensive thermal and hydraulic analysis is needed. Compact heat exchangers require very small flow channels and fins to achieve high heat transfer rates and thermal effectiveness. However, studying such small features computationally contributes little to the understanding of component level phenomena and requires prohibitive computational effort using computational fluid dynamics (CFD). To address this issue, the analysis developed here uses an effective porous media (EPM) approach; this greatly reduces the computation time and produces results with the appropriate resolution [1]. This EPM fluid dynamics and heat transfer computational code has been named the Compact Heat Exchanger Explicit Thermal and Hydraulics (CHEETAH) code. CHEETAH solves for the two-dimensional steady-state and transient temperature and flow distributions in the IHX including the complicating effects of temperature-dependent fluid thermo-physical properties. Temperature- and pressure-dependent fluid properties are evaluated by CHEETAH and the thermal effectiveness of the IHX is also calculated. Furthermore, the temperature distribution can then be imported into a finite element analysis (FEA) code for mechanical stress analysis using the EPM methods developed earlier by the University of California, Berkeley, for global and local stress analysis [2]. These simulation tools will also allow the heat exchanger design to be improved through an

  2. 40 CFR 90.329 - Catalyst thermal stress test.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 20 2010-07-01 2010-07-01 false Catalyst thermal stress test. 90.329... Equipment Provisions § 90.329 Catalyst thermal stress test. (a) Oven characteristics. The oven used for thermally stressing the test catalyst must be capable of maintaining a temperature of 500 ±5 °C and 1000 ±10...

  3. Assessment study of RELAP5/SCDAP capability to reproduce TALL facility thermal hydraulic behavior

    Energy Technology Data Exchange (ETDEWEB)

    Fiori, F., E-mail: filippofiori85@gmail.com [INET, Tsinghua University, Beijing 100084 (China); Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China); Zhou, Z.W. [INET, Tsinghua University, Beijing 100084 (China); Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2015-12-15

    Highlights: • We modified the RELAP5/SCDAP code to work with LBE and glycerol. • A RELAP5 model has been set up following preset choices and guidelines. • Seven different transients have been simulated. • Two different HTC correlation have been studied. - Abstract: The paper presents the assessment of RELAP5/SCDAP code capabilities to simulate the thermal–hydraulic behavior of liquid metal. The code has been recently modified to work with liquid metal; new heat transfer correlations have been implemented. As the code is widely used in our institute during the design process of the Chinese ADS reactor the assessment of the newly modified RELAP5/SCDAP is seen as a necessary step to ensure the quality of the code results. The present paper focuses on the simulation of the transients performed on the TALL facility. TALL has been constructed and operated at KTH Royal Institute of Technology of Stockholm. The full height facility was designed and operated to investigate the heat transfer performance of different heat exchangers and the thermal–hydraulic characteristics of natural and forced circulation flow under steady and transient conditions. Two different configurations are available for the TALL facility however only one is simulated for the present study with seven transients analyzed. Different LBE heat transfer correlations are compared for the calculations. A consistent and systematic approach for the nodalization development and assessment procedures that respond to the IAEA guidelines is discussed and thoroughly applied. The procedures and the database developed constitute the base in our institute for further study when more experimental data is made available.

  4. Thermal-Structures and Materials Testing Laboratory

    Science.gov (United States)

    Teate, Anthony A.

    1997-01-01

    Since its inception and successful implementation in 1997 at James Madison University, the Thermal Structures and Materials Testing Laboratory (T-SaMTL) funded by the NASA Langley Research Center is evolving into one of the University's premier and exemplary efforts to increase minority representation in the sciences and mathematics. Serving ten (10) students and faculty directly and almost fifty (50) students indirectly, T-SAMTL, through its recruitment efforts, workshops, mentoring program, tutorial services and its research and computational laboratories has marked the completion of the first year with support from NASA totaling $ 100,000. Beginning as an innovative academic research and mentoring program for underrepresented minority science and mathematics students, the program now boasts a constituency which consists of 50% graduating seniors in the spring of 1998 with 50% planning to go to graduate school. The program's intent is to increase the number of underrepresented minorities who receive doctoral degrees in the sciences by initiating an academically enriched research program aimed at strengthening the academic and self actualization skills of undergraduate students with the potential to pursue doctoral study in the sciences. The program provides financial assistance, academic enrichment, and professional and personal development support for minority students who demonstrate the potential and strong desire to pursue careers in the sciences and mathematics. James Madison University was awarded the first $100,000, in April 1997, by The NASA Langley Research Center for establishment and support of its Thermal Structures and Materials Testing

  5. Testing hypotheses that link wood anatomy to cavitation resistance and hydraulic conductivity in the genus Acer.

    Science.gov (United States)

    Lens, Frederic; Sperry, John S; Christman, Mairgareth A; Choat, Brendan; Rabaey, David; Jansen, Steven

    2011-05-01

    • Vulnerability to cavitation and conductive efficiency depend on xylem anatomy. We tested a large range of structure-function hypotheses, some for the first time, within a single genus to minimize phylogenetic 'noise' and maximize detection of functionally relevant variation. • This integrative study combined in-depth anatomical observations using light, scanning and transmission electron microscopy of seven Acer taxa, and compared these observations with empirical measures of xylem hydraulics. • Our results reveal a 2 MPa range in species' mean cavitation pressure (MCP). MCP was strongly correlated with intervessel pit structure (membrane thickness and porosity, chamber depth), weakly correlated with pit number per vessel, and not related to pit area per vessel. At the tissue level, there was a strong correlation between MCP and mechanical strength parameters, and some of the first evidence is provided for the functional significance of vessel grouping and thickenings on inner vessel walls. In addition, a strong trade-off was observed between xylem-specific conductivity and MCP. Vessel length and intervessel wall characteristics were implicated in this safety-efficiency trade-off. • Cavitation resistance and hydraulic conductivity in Acer appear to be controlled by a very complex interaction between tissue, vessel network and pit characteristics. © 2010 The Authors. New Phytologist © 2010 New Phytologist Trust.

  6. Hydraulic characterisation of carbonate aquifers with pumping tests - comparison of analytical and numerical data evaluation methods

    Science.gov (United States)

    Baierl, M.; Kordilla, J.; Reimann, T.; Dörfliger, N.; Sauter, M.; Geyer, T.

    2012-04-01

    This work deals with the analysis of pumping tests in strongly heterogeneous media. Pumping tests were performed in the catchment area of the Lez spring (South of France), which is composed of carbonate rocks. Pumping rates for the different tests varied between 0.04 l/s - 0.7 l/s, i.e. the radius of influence of the cone of depression is small. The investigated boreholes are characterised by tight rocks, moderate fractures and karstified zones. The observed drawdown curves are clearly influenced by the rock characteristics. Single drawdown curves show S-shape character. Data evaluation was performed with the solution approaches of Theis (1935) and Gringarten-Ramey (1974), which are implemented in the employed software AQTESOLV (Pro 4.0). Parameters were varied in reliable data ranges with consideration of reported values in the literature. The Theis method analyses unsteady flow in homogeneous confined aquifers. The Gringarten-Ramey solution describes the drawdown in a well connected to a single horizontal fracture. The Theis curve fails to represent the characteristics for nearly all of the measured drawdown curves, while the Gringarten-Ramey method shows moderate graphical fits with a small residual sum of squares between fitted and observed drawdown curves. This highlights the importance of heterogeneities in the hydraulic parameter field at local scale. The determined hydraulic conductivities of the rock are in reasonable ranges varying between 1E-04 m/s and 1E-08 m/s. Wellbore skin effects need to be discussed further in detail. While the analytical solutions are only valid for specific geometrical and hydraulic configurations, numerical models can be applied to simulate pumping tests in complex heterogeneous media with different boundary conditions. For that reason, a two dimensional, axisymmetric numerical model, using COMSOL (Multiphysics 4.1), is set up. In a first step, the model is validated with the simulated curves from the analytical solutions under

  7. Tests Performed on Hydraulic Turbines at Commissioning or after Capital Repairs. Part II. Tests Performed on a 6.5 MW Kaplan Turbine

    Directory of Open Access Journals (Sweden)

    Adrian Cuzmoş

    2015-07-01

    Full Text Available The paper presents the tests performed on a hydraulic turbine on commissioning, the devices, test methods and the results obtained from the respective tests, as well as the conclusions and recommendations resulted from these tests. This kind of tests can be performed for the verification of guarantees.

  8. Neutronic and thermal-hydraulic analysis of new irradiation channels inside the Moroccan TRIGA Mark II research reactor core.

    Science.gov (United States)

    Chham, E; El Bardouni, T; Benaalilou, K; Boukhal, H; El Bakkari, B; Boulaich, Y; El Younoussi, C; Nacir, B

    2016-10-01

    This study was conducted to improve the capacity of radioisotope production in the Moroccan TRIGA Mark II research reactor, which is considered as one of the most important applications of research reactors. The aim of this study is to enhance the utilization of TRIGA core in the field of neutron activation and ensure an economic use of the fuel. The main idea was to create an additional irradiation channel (IC) inside the core. For this purpose, three new core configurations are proposed, which differ according to the IC position in the core. Thermal neutron flux distribution and other neutronic safety parameters such as power peaking factors, excess reactivity, and control rods worth reactivity were calculated using the Monte Carlo N-Particle Transport (MCNP) code and neutron cross-section library based on ENDF/B-VII evaluation. The calculated thermal flux in the central thimble (CT) and in the added IC for the reconfigured core is compared with the thermal flux in the CT of the existing core, which is taken as a reference. The results show that all the obtained fluxes in CTs are very close to the reference value, while a remarkable difference is observed between the fluxes in the new ICs and reference. This difference depends on the position of IC in the reactor core. To demonstrate that the Moroccan TRIGA reactor could safely operate at 2MW, with new configurations based on new ICs, different safety-related thermal-hydraulic parameters were investigated. The PARET model was used in this study to verify whether the safety margins are met despite the new modifications of the core. The results show that it is possible to introduce new ICs safely in the reactor core, because the obtained values of the parameters are largely far from compromising the safety of the reactor. Copyright © 2016 Elsevier Ltd. All rights reserved.

  9. Variations in hydraulic conductivity with scale of measurement during aquifer tests in heterogeneous, porous carbonate rocks

    Science.gov (United States)

    Schulze-Makuch, Dirk; Cherkauer, Douglas S.

    Previous studies have shown that hydraulic conductivity of an aquifer seems to increase as the portion of the aquifer tested increases. To date, such studies have all relied on different methods to determine hydraulic conductivity at each scale of interest, which raises the possibility that the observed increase in hydraulic conductivity is due to the measurement method, not to the scale. This study analyzes hydraulic conductivity with respect to scale during individual aquifer tests in porous, heterogeneous carbonate rocks in southeastern Wisconsin, USA. Results from this study indicate that hydraulic conductivity generally increases during an individual test as the volume of aquifer impacted increases, and the rate of this increase is the same as the rate of increase determined by using different measurement methods. Thus, scale dependence of hydraulic conductivity during single tests does not depend on the method of measurement. This conclusion is supported by 22 of 26 aquifer tests conducted in porous-flow-dominated carbonate units within the aquifer. Instead, scale dependency is probably caused by heterogeneities within the aquifer, a conclusion supported by digital simulation. All of the observed types of hydraulic-conductivity variations with scale during individual aquifer tests can be explained by a conceptual model of a simple heterogeneous aquifer composed of high-conductivity zones within a low-conductivity matrix. Résumé Certaines études ont montré que la conductivité hydraulique d'un aquifère semble augmenter en même temps que la partie testée de l'aquifère s'étend. Jusqu'à présent, ces études ont toutes reposé sur des méthodes de détermination de la conductivité hydraulique différentes pour chaque niveau d'échelle, ce qui a conduit à penser que l'augmentation observée de la conductivité hydraulique pouvait être due aux méthodes de mesure et non à l'effet d'échelle. Cette étude analyse la conductivité hydraulique par

  10. Spatial variation in hydraulic conductivity determined by slug tests in the Canadian River alluvium near the Norman Landfill, Norman, Oklahoma

    Science.gov (United States)

    Scholl, Martha A.; Christenson, Scott C.

    1998-01-01

    Slug tests were used to characterize hydraulic conductivity variations at a spatial scale on the order of meters in the alluvial aquifer downgradient of the Norman Landfill. Forty hydraulic conductivity measurements were made, most along a 215-meter flow path transect. Measured hydraulic conductivity, excluding clayey layers, ranged from 8.4 ? 10-7 to 2.8 ? 10-4 meters per second, with a median value of 6.6 ? 10-5 meters per second. The hydraulic conductivity measurements yield a preliminary concept of the permeability structure of the aquifer along this transect. A low hydraulic conductivity silt-clay layer at about 4 meters below the water table and a high hydraulic conductivity layer at the base of the aquifer appear to have the most potential to affect contaminant transport. Specific conductance measurements show the leachate plume along this transect becomes attenuated between 150 and 200 meters downgradient of the landfill, except at the base of the aquifer, where it extends at least 225 meters downgradient of the landfill.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

    {sub th} power and electricity generation with 100 kW{sub th} idle power. Consequently, KANUTER has the characteristics of a compact and lightweight system, excellent propellant efficiency, bimodal capability, and mission versatility as indicated in the reference design parameters. This thermo-hydraulic design analysis was carried out to estimate the optimum FWT of the unique SLHC fuel design in the core and thereby the maximum rocket performance. The FWT affects the mechanical strength of the SLHC fuel assembly as well as the thermo-hydraulic capability mainly depending on the heat transfer area of fuel. The thicker fuel wafer is mechanically strong with low pressure drop, while the thinner fuel wafer is thermally robust with less mechanical strength and higher shear stress in the core.

  12. Can one identify karst conduit networks geometry and properties from hydraulic and tracer test data?

    Science.gov (United States)

    Borghi, Andrea; Renard, Philippe; Cornaton, Fabien

    2016-04-01

    Karst aquifers are characterized by extreme heterogeneity due to the presence of karst conduits embedded in a fractured matrix having a much lower hydraulic conductivity. The resulting contrast in the physical properties of the system implies that the system reacts very rapidly to some changes in the boundary conditions and that numerical models are extremely sensitive to small modifications in properties or positions of the conduits. Furthermore, one major issue in all those models is that the location and size of the conduits is generally unknown. For all those reasons, estimating karst network geometry and their properties by solving an inverse problem is a particularly difficult problem. In this paper, two numerical experiments are described. In the first one, 18,000 flow and transport simulations have been computed and used in a systematic manner to assess statistically if one can retrieve the parameters of a model (geometry and radius of the conduits, hydraulic conductivity of the conduits) from head and tracer data. When two tracer test data sets are available, the solution of the inverse problems indicate with high certainty that there are indeed two conduits and not more. The radius of the conduits are usually well identified but not the properties of the matrix. If more conduits are present in the system, but only two tracer test data sets are available, the inverse problem is still able to identify the true solution as the most probable but it also indicates that the data are insufficient to conclude with high certainty. In the second experiment, a more complex model (including non linear flow equations in conduits) is considered. In this example, gradient-based optimization techniques are proved to be efficient for estimating the radius of the conduits and the hydraulic conductivity of the matrix in a promising and efficient manner. These results suggest that, despite the numerical difficulties, inverse methods should be used to constrain numerical

  13. Flexible parallel implicit modelling of coupled thermal-hydraulic-mechanical processes in fractured rocks

    Science.gov (United States)

    Cacace, Mauro; Jacquey, Antoine B.

    2017-09-01

    Theory and numerical implementation describing groundwater flow and the transport of heat and solute mass in fully saturated fractured rocks with elasto-plastic mechanical feedbacks are developed. In our formulation, fractures are considered as being of lower dimension than the hosting deformable porous rock and we consider their hydraulic and mechanical apertures as scaling parameters to ensure continuous exchange of fluid mass and energy within the fracture-solid matrix system. The coupled system of equations is implemented in a new simulator code that makes use of a Galerkin finite-element technique. The code builds on a flexible, object-oriented numerical framework (MOOSE, Multiphysics Object Oriented Simulation Environment) which provides an extensive scalable parallel and implicit coupling to solve for the multiphysics problem. The governing equations of groundwater flow, heat and mass transport, and rock deformation are solved in a weak sense (either by classical Newton-Raphson or by free Jacobian inexact Newton-Krylow schemes) on an underlying unstructured mesh. Nonlinear feedbacks among the active processes are enforced by considering evolving fluid and rock properties depending on the thermo-hydro-mechanical state of the system and the local structure, i.e. degree of connectivity, of the fracture system. A suite of applications is presented to illustrate the flexibility and capability of the new simulator to address problems of increasing complexity and occurring at different spatial (from centimetres to tens of kilometres) and temporal scales (from minutes to hundreds of years).

  14. Thermal-hydraulics analysis of a PWR reactor using zircaloy and carbide silicon reinforced with type S fibers as fuel claddings: Simulation of a channel blockage transient

    Energy Technology Data Exchange (ETDEWEB)

    Matuck, Vinicius; Ramos, Mario C.; Faria, Rochkhudson B.; Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia, E-mail: rochkdefaria@yahoo.com.br, E-mail: matuck747@gmail.com, E-mail: patricialire@yahoo.com.br, E-mail: marc5663@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    A detailed thermal-hydraulic reactor model using as reference data from the Angra 2 Final Safety Analysis Report (FSAR) has been developed and SiC reinforced with Hi-Nicalon type S fibers (SiC HNS) was used as fuel cladding. The goal is to compare its behavior from the thermal viewpoint with the Zircaloy, at the steady- state and transient conditions. The RELAP-3D was used to perform the thermal-hydraulic analysis and a blockage transient has been investigated at full power operation. The transient considered is related to total obstruction of a core cooling channel of one fuel assembly. The calculations were performed using a point kinetic model. The reactor behavior after this transient was analyzed and the time evolution of cladding and coolant temperatures mass flow and void fraction are presented. (author)

  15. Model of a nuclear thermal test pipe using ATHENA. Master's thesis

    Energy Technology Data Exchange (ETDEWEB)

    Dibben, M.J.

    1992-03-01

    Nuclear thermal propulsion offers significant improvements in rocket engine specific impulse over rockets employing chemical propulsion. The computer code ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) was used in a parametric analysis of fuelpipe. The fuelpipe is an annular particle bed fuel element of the reactor with radially inward flow of hydrogen through it. The outlet temperature of the hydrogen is parametrically related to key effects, including the effect of reactor power at two different pressure drops, the effect of the power coupling factor of the Annular Core Research Reactor, and the effect of hydrogen flow. Results show that the outlet temperature is linearly related to the reactor power and nonlinearly to the change in pressure drop. The linear relationship at higher temperatures is probably not valid due to dissociation of hydrogen. Once thermal properties of hydrogen become available, the ATHENA model for this study could easily be modified to test this conjecture.

  16. A test of the hydraulic vulnerability segmentation hypothesis in angiosperm and conifer tree species.

    Science.gov (United States)

    Johnson, Daniel M; Wortemann, Remi; McCulloh, Katherine A; Jordan-Meille, Lionel; Ward, Eric; Warren, Jeffrey M; Palmroth, Sari; Domec, Jean-Christophe

    2016-08-01

    Water transport from soils to the atmosphere is critical for plant growth and survival. However, we have a limited understanding about many portions of the whole-tree hydraulic pathway, because the vast majority of published information is on terminal branches. Our understanding of mature tree trunk hydraulic physiology, in particular, is limited. The hydraulic vulnerability segmentation hypothesis (HVSH) stipulates that distal portions of the plant (leaves, branches and roots) should be more vulnerable to embolism than trunks, which are nonredundant organs that require a massive carbon investment. In the current study, we compared vulnerability to loss of hydraulic function, leaf and xylem water potentials and the resulting hydraulic safety margins (in relation to the water potential causing 50% loss of hydraulic conductivity) in leaves, branches, trunks and roots of four angiosperms and four conifer tree species. Across all species, our results supported strongly the HVSH as leaves and roots were less resistant to embolism than branches or trunks. However, branches were consistently more resistant to embolism than any other portion of the plant, including trunks. Also, calculated whole-tree vulnerability to hydraulic dysfunction was much greater than vulnerability in branches. This was due to hydraulic dysfunction in roots and leaves at less negative water potentials than those causing branch or trunk dysfunction. Leaves and roots had narrow or negative hydraulic safety margins, but trunks and branches maintained positive safety margins. By using branch-based hydraulic information as a proxy for entire plants, much research has potentially overestimated embolism resistance, and possibly drought tolerance, for many species. This study highlights the necessity to reconsider past conclusions made about plant resistance to drought based on branch xylem only. This study also highlights the necessity for more research of whole-plant hydraulic physiology to better

  17. Flexible thermal cycle test equipment for concentrator solar cells

    Science.gov (United States)

    Hebert, Peter H [Glendale, CA; Brandt, Randolph J [Palmdale, CA

    2012-06-19

    A system and method for performing thermal stress testing of photovoltaic solar cells is presented. The system and method allows rapid testing of photovoltaic solar cells under controllable thermal conditions. The system and method presents a means of rapidly applying thermal stresses to one or more photovoltaic solar cells in a consistent and repeatable manner.

  18. Nuclear Thermal Propulsion Ground Test History

    Science.gov (United States)

    Gerrish, Harold P.

    2014-01-01

    Nuclear Thermal Propulsion (NTP) was started in 1955 under the Atomic Energy Commission as project Rover and was assigned to Los Alamos National Laboratory. The Nevada Test Site was selected in 1956 and facility construction began in 1957. The KIWI-A was tested on July 1, 1959 for 5 minutes at 70MW. KIWI-A1 was tested on July 8, 1960 for 6 minutes at 85MW. KIWI-A3 was tested on October 10, 1960 for 5 minutes at 100MW. The National Aeronautics and Space Administration (NASA) was formed in 1958. On August 31, 1960 the AEC and NASA established the Space Nuclear Propulsion Office and named Harold Finger as Director. Immediately following the formation of SNPO, contracts were awarded for the Reactor In Flight Test (RIFT), master plan for the Nuclear Rocket Engine Development Station (NRDS), and the Nuclear Engine for Rocket Vehicle Application (NERVA). From December 7, 1961 to November 30, 1962, the KIWI-B1A, KIWI-B1B, and KIWI-B4A were tested at test cell A. The last two engines were only tested for several seconds before noticeable failure of the fuel elements. Harold Finger called a stop to any further hot fire testing until the problem was well understood. The KIWI-B4A cold flow test showed the problem to be related to fluid dynamics of hydrogen interstitial flow causing fuel element vibrations. President Kennedy visited the NTS one week after the KIWI-B4A failure and got to see the engine starting to be disassembled in the maintenance facility. The KIWI-B4D and KIWI-B4E were modified to not have the vibration problems and were tested in test cell C. The NERVA NRX program started testing in early 1964 with NRX-A1 cold flow test series (unfueled graphite core), NRX-A2 and NRX-A3 power test series up to 1122 MW for 13 minutes. In March 1966, the NRX-EST (Engine System Test) was the first breadboard using flight functional relationship and total operating time of 116 minutes. The NRX-EST demonstrated the feasibility of a hot bleed cycle. The NRX-A5 had multiple start

  19. Thermal-Hydraulic Effect of Pattern of Wire-wrap Spacer in 19-pin Rod Bundle for SFR Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yeong Shin; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Park, Seong Dae [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    As sodium-cooled fast reactor (SFR) has been considered the most promising reactor type for future and prototype gen-IV SFR has been developed actively in Korea, thermal-hydraulic aspects of the SFR fuel assembly have the important role for the reactor safety analysis. In PGSFR fuel assembly, 271 pins of fuel rods are tightly packed in triangular array inside hexagonal duct, and wire is wound helically per each fuel rod with regular pattern to assure the gap between rods and prevent the collision, which is called wire-wrapped spacer. Due to helical shape of the wire-wrapped spacer, flow inside duct can have stronger turbulent characteristics and thermal mixing effect. However, many studies showed the possible wake from swirl flow inside subchannel, which cause local hot spot. To prevent the wake flow and improve thermal mixing, new pattern of wire wrap spacer was suggested. To evaluate the effect of wire wrap spacer pattern, CFD analysis was performed for 19-pin rod bundle with comparison of conventional and U-pattern wire wrap spacer. To prevent the wake due to same direction of swirl flow, 7-rod unit pattern of wire spacer, which are arranged to have different rotational direction of wire with adjacent rods and center rod without wire wrap was proposed. From simulation results, swirl flow across gap conflicts its rotation direction causing wake flow from the regular pattern of the conventional one, which generates local hot spot near cladding. With U-pattern of wire wrap spacer, heat transfer in subchannel can be enhanced with evenly distributed cross flow without compensating pressure loss. From the results, the pattern of wire wrap spacer can influence the both heat transfer characteristics and pressure drop, with flow structures generated by wire wrap spacer.

  20. The R&D PERFROI Project on Thermal Mechanical and Thermal Hydraulics Behaviors of a Fuel Rod Assembly during a Loss of Coolant Accident

    Energy Technology Data Exchange (ETDEWEB)

    Repetto, G. [Institut de Radioprotection et de Surete Nucleaire, Cadarache (France); Dominguez, C. [Institut de Radioprotection et de Surete Nucleaire, Cadarache (France); Durville, B. [Institut de Radioprotection et de Surete Nucleaire, Cadarache (France); Carnemolla, S. [Institut de Radioprotection et de Surete Nucleaire, Cadarache (France); Campello, D. [Institut National des Sciences Appliques, Lyon (France); Tardiff, N. [Institut National des Sciences Appliques, Lyon (France); Gradeck, M. [Univ. de Lorraine, Nancy, France. LEMTA

    2015-09-04

    The safety principle in case of a LOCA is to preserve the short and long term coolability of the core. The associated safety requirements are to ensure the resistance of the fuel rods upon quench and post-quench loads and to maintain a coolable geometry in the core. An R&D program has been launched by IRSN with the support of EDF, to perform both experimental and modeling activities in the frame of the LOCA transient, on technical issues such as: - flow blockage within a fuel rods bundle and its potential impact on coolability, - fuel fragment relocation in the ballooned areas: its potential impact on cladding PCT (Peak Cladding Temperature) and on the maximum oxidation rate, - potential loss of cladding integrity upon quench and post-quench loads. The PERFROI project (2014-2019) focusing on the first above issue, is structured in two axes: 1. axis 1: thermal mechanical behavior of deformation and rupture of cladding taking into account the contact between fuel rods; specific research at LaMCoS laboratory focus on the hydrogen behavior in cladding alloys and its impact on the mechanical behavior of the rod; and, 2. axis 2: thermal hydraulics study of a partially blocked region of the core (ballooned area taking into account the fuel relocation with local over power), during cooling phase by water injection; More detailed activities foreseen in collaboration with LEMTA laboratory will focus on the characterization of two phase flows with heat transfer in deformed structures.

  1. Development of a thermal hydraulic modelling of ground water of the Malm in the Munich metropolitan area; Entwicklung einer thermisch-hydraulischen Grundwassermodellierung des Malm im Grossraum Muenchen

    Energy Technology Data Exchange (ETDEWEB)

    Dussel, M.; Lueschen, E.; Thomas, R. [Leibniz-Institut fuer Angewandte Geophysik, Hannover (DE)] (and others)

    2011-10-24

    The mutual potential influence of geothermal duplicates and the scientific investigation of the relationship between seismic and hydraulic parameters are investigated in the joint research project 'Geothermal characterization of fractured karst limestone aquifers in the Munich metropolitan area'. Thirteen doublets and triplets being in production or sunk illustrate the great geothermal potential and provide important data on the development of a thermal-hydraulic modeling of the reservoir. 3D seismic Unterhaching, 3D structural model, hydrogeological model and a high-resolution 3D temperature model form the basis of the numerical modeling. Different seismic signatures, seismic attributes and variations in the interval velocities characterize the ground geophysically, and were interpreted under consideration of geological and hydrogeological background information as well as borehole measurements in terms of hydraulically conductive homogeneous areas. For the Munich metropolitan area, the numerical modeling is a decision aid for the future optimized and sustainable hydrothermal utilization of the Malm aquifer.

  2. Supercritical Water Nuclear Steam Supply System: Innovations In Materials, Neutronics & Thermal-Hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Mark Anderson; M.L. Corradini; K. Sridharan; P. WIlson; D. Cho; T.K. Kim; S. Lomperski

    2004-09-02

    In the 1990's supercritical light-water reactors were considered in conceptual designs. A nuclear reactor cooled by supercritical waster would have a much higher thermal efficiency with a once-through direct power cycle, and could be based on standardized water reactor components (light water or heavy water). The theoretical efficiency could be improved by more than 33% over that of other water reactors and could be simplified with higher reliability; e.g., a boiling water reactor without steam separators or dryers.

  3. A thermal-hydraulic drift-flux based mixture-fluid model for the description of single- and two-phase flow along a general coolant channel

    Energy Technology Data Exchange (ETDEWEB)

    Alois Hoeld [Bernaysstr. 16A, D-80937 Munich (Germany)

    2005-07-01

    constructed with the aim to relate the mean nodal and nodal boundary function values appearing in the nodalized differential eqs. and to yield necessary gradients at SC boundaries. It represents together with the very thoroughly tested packages for drift flux, heat transfer and single- and two-phase friction factors the central part of the theoretical model and module CCM and guarantees for the exact solution of the mixture-fluid eqs. The module has been derived on the basis of experiences of many years work with the development of an effective theoretical model and corresponding non-linear one-dimensional digital code UTSG-2 for vertical, natural-circulation U-tube steam generators as being used in PWR nuclear power plants. The general module CCM has now been used to replace, in a first step, the special coolant channel elements of the thoroughly tested U-tube steam generator code UTSG-2, constructing thus (among other improvements and additions) the advanced code version UTSG-3. Some of the most characteristic previous calculations with the UTSG-2 code could then be taken as benchmark cases. Post-calculating them showed good agreement, presenting a very convincing argument for the validation of the UTSG-3 code and thus also the underlying code package CCM. This universally applicable thermal-hydraulic code package CCM can and has been taken as a basic element for the simulation of the thermal-hydraulic situation of different complex systems (different types of steam generators, 3D nuclear reactor cores calculating the mass flow distribution into different parallel channels after non-symmetric perturbations), yielding very satisfactory results. The 3-eq. ('separate-region') mixture fluid model can be seen as a counterpart to the at present very much preferred 'separate-phase models' where each phase is treated separately, i.e., the 3 conservation eqs. are splitted into 4, 5 or even 6 field equations, connected by exchange (=closure) terms which represent

  4. Interpretation of Flow Logs from Nevada Test Site Boreholes to Estimate Hydraulic conductivity Using Numerical Simulations Constrained by Single-Well Aquifer Tests

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, C. Amanda; Halford, Keith J.; Laczniak, Randell J.

    2010-02-12

    Hydraulic conductivities of volcanic and carbonate lithologic units at the Nevada Test Site were estimated from flow logs and aquifer-test data. Borehole flow and drawdown were integrated and interpreted using a radial, axisymmetric flow model, AnalyzeHOLE. This integrated approach is used because complex well completions and heterogeneous aquifers and confining units produce vertical flow in the annular space and aquifers adjacent to the wellbore. AnalyzeHOLE simulates vertical flow, in addition to horizontal flow, which accounts for converging flow toward screen ends and diverging flow toward transmissive intervals. Simulated aquifers and confining units uniformly are subdivided by depth into intervals in which the hydraulic conductivity is estimated with the Parameter ESTimation (PEST) software. Between 50 and 150 hydraulic-conductivity parameters were estimated by minimizing weighted differences between simulated and measured flow and drawdown. Transmissivity estimates from single-well or multiple-well aquifer tests were used to constrain estimates of hydraulic conductivity. The distribution of hydraulic conductivity within each lithology had a minimum variance because estimates were constrained with Tikhonov regularization. AnalyzeHOLE simulated hydraulic-conductivity estimates for lithologic units across screened and cased intervals are as much as 100 times less than those estimated using proportional flow-log analyses applied across screened intervals only. Smaller estimates of hydraulic conductivity for individual lithologic units are simulated because sections of the unit behind cased intervals of the wellbore are not assumed to be impermeable, and therefore, can contribute flow to the wellbore. Simulated hydraulic-conductivity estimates vary by more than three orders of magnitude across a lithologic unit, indicating a high degree of heterogeneity in volcanic and carbonate-rock units. The higher water transmitting potential of carbonate-rock units relative

  5. Thermal and hydraulic considerations regarding the fate of water discharged by the outflow channels to the Martian northern plains

    Science.gov (United States)

    Clifford, S. M.

    1993-01-01

    The identification of possible shorelines in the Martian northern plains suggests that the water discharged by the circum-Chryse outflow channels may have led to the formation of transient seas, or possibly even an ocean, covering as much as one-third of the planet. Speculations regarding the possible fate of this water have included local ponding and reinfiltration into the crust; freezing, sublimation, and eventual cold-trapping at higher latitudes; or the in situ survival of this now frozen water to the present day -- perhaps aided by burial beneath a protective cover of eolian sediment or lavas. Although neither cold-trapping at higher latitudes nor the subsequent freezing and burial of flood waters can be ruled out, thermal and hydraulic considerations effectively eliminate the possibility that any significant reassimilation of this water by local infiltration has occurred given climatic conditions resembling those of today. The arguments against the local infiltration of flood water into the northern plains are two-fold. First, given the climatic and geothermal conditions that are thought to have prevailed on Mars during the Late Hesperian (the period of peak outflow channel activity in the northern plains), the thickness of the cryosphere in Chryse Planitia is likely to have exceeded 1 km. A necessary precondition for the widespread occurrence of groundwater is that the thermodynamic sink represented by the cryosphere must already be saturated with ice. For this reason, the ice-saturated cryosphere acts as an impermeable barrier that effectively precludes the local resupply of subpermafrost groundwater by the infiltration of water discharged to the surface by catastraphic floods. Note that the problem of local infiltration is not significantly improved even if the cryosphere were initially dry, for as water attempts to infiltrate the cold, dry crust, it will quickly freeze, creating a seal that prevents any further infiltration from the ponded water above

  6. Quantitative Stress Determination by Wire-line Hydraulic Fracturing Tests in the ANDRILL South McMurdo Sound Drillhole

    Science.gov (United States)

    Schmitt, D. R.; Wilson, T. J.; Pierdominici, S.; Jarrard, R. D.; Paulsen, T. S.; Wonik, T.; Handwerger, D.

    2008-12-01

    In general, knowledge of the state of stress within the Antarctic lithosphere remains largely unconstrained due in part to Antarctica's inaccessibility and because of the paucity of seismic focal mechanism solutions. As such, an important component of the ANDRILL project was to acquire new information on the stress directions and magnitudes within the upper crust. Extensive fracture mapping of the core and oriented ultrasonic televiewer logging, as presented in a related contribution, provided strong constraints on both stress orientations and the faulting regime. The first hydraulic fracturing stress determinations in Antarctica were also carried out to provide complimentary stress magnitudes. These measurements were accomplished using a wireline transported hydraulic fracturing system consisting, essentially, of two synthetic polymer packers that were inflated to isolate the 1-m long pressurization interval. The tests were conducted at the end of the drilling and logging in a ~130-m section specially drilled below 1000 mbsf for hydraulic fractures. This zone was drilled primarily through competent, dense, and low permeability diamicts. The core fracture and televiewer logging information was used to site twenty separate measurements. Classic hydraulic fracturing pressurization records with unambiguous breakdown and fracture closure pressures were obtained in about half of the measurements. The remaining tests provided fracture propagation and closure pressures that are consistent with the classic breakdown curves, but they did not include a clear breakdown pressure and are indicative of the reopening of either natural or drilling induced fractures in the wellbore wall. Comparison of the televiewer images obtained before and after the hydraulic fracturing tests highlighted the existence of at least one artificial hydraulic fracture; more could not be obtained because blockage of the wellbore did not allow the lowest sections to be logged a second time. The

  7. Analyses of Instability Events in the Peach Bottom-2 BWR Using Thermal-Hydraulic and 3D Neutron Kinetic Coupled Codes Technique

    Directory of Open Access Journals (Sweden)

    Antonella Lombardi Costa

    2008-01-01

    Full Text Available Boiling water reactor (BWR instabilities may occur when, starting from a stable operating condition, changes in system parameters bring the reactor towards an unstable region. In order to design more stable and safer core configurations, experimental and theoretical studies about BWR stability have been performed to characterise the phenomenon and to predict the conditions for its occurrence. In this work, contributions to the study of BWR instability phenomena are presented. The RELAP5/MOD3.3 thermal-hydraulic (TH system code and the PARCS-2.4 3D neutron kinetic (NK code were coupled to simulate BWR transients. Different algorithms were used to calculate the decay ratio (DR and the natural frequency (NF from the power oscillation predicted by the transient calculations as two typical parameters used to provide a quantitative description of instabilities. The validation of the code model set up for the Peach Bottom Unit 2 BWR plant is performed against low-flow stability tests (LFSTs. The four series of LFST have been performed during the first quarter of 1977 at the end of cycle 2 in Pennsylvania. The tests were intended to measure the reactor core stability margins at the limiting conditions used in design and safety analyses.

  8. Hydraulic machine tests for compression of a quasi-brittle material at medium strain rate

    Science.gov (United States)

    Quirion, Y.; Lesaffre, A. S.

    2006-08-01

    This paper describes an experimental device used to determine the dynamic compressive behaviour of quasi-brittle material at medium strain rates (1 to 100 s - 1). The tool combines a servo-hydraulic machine with a high-speed photography. Tests consist in compressing a sample between a dynamic jack and an instrumented anvil according to the direct impact method. The main difficulty of brittle material testing is to achieve dynamic equilibrium in the sample before failure because of their low failure strains. Furthermore, oscillations phenomena disturb load measurement. In this paper, we present adequate methods in order to carry out homogeneous testing and to simplify data interpretation. Two experimental configurations are developed. We use firstly the anvil as a load cell for low impact velocity and secondly the wave propagation in the anvil for medium impact velocity. Finally, in order to investigate experimentally the strain uniformity, axial strain measurements are quantified by image processing. Results are compared with experimental ones obtained on a crossbow system.

  9. FEM (finite element method) thermal modeling and thermal hydraulic performance of an enhanced thermal conductivity UO2/BeO composite fuel

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Wenzhong [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Purdue Univ., West Lafayette, IN (United States)

    2011-03-24

    An enhanced thermal conductivity UO2-BeO composite nuclear fuel was studied. A methodology to generate ANSYS (an engineering simulation software) FEM (Finite Element Method) thermal models of enhanced thermal conductivity oxide nuclear fuels was developed. The results showed significant increase in the fuel thermal conductivities and have good agreement with the measured ones. The reactor performance analysis showed that the decrease in centerline temperature was 250-350K for the UO2-BeO composite fuel, and thus we can improve nuclear reactors' performance and safety, and high-level radioactive waste generation.

  10. Evaluation of Hydraulic Loads on the Runner Blades of a Kaplan Turbine using CFD Simulation and Model Test

    Directory of Open Access Journals (Sweden)

    Zoltan-Iosif Korka

    2016-10-01

    Full Text Available CFD (Computational Fluid Dynamic is today a standard procedure for analyzing and simulating the flow through several hydraulic machines. In this process, the fluid flow domain is divided into small volumes where the governing equations are converted into algebraic ones, which are numerically solved. Computational results strongly depend on the applied mathematical model and on the numerical methods used for converting the governing equations into the algebraic ones. The goal of the paper is to evaluate, by numerical simulation, the hydraulic loads (forces and torques on the runner blades of an existent Kaplan turbine and to compare them with the experimental results obtained from model test.

  11. Thermal-hydraulic analysis of the HL-2M divertor using an homogeneous equilibrium model

    Science.gov (United States)

    Lu, Yong; Cai, Lijun; Liu, Yuxiang; Liu, Jian; Yuan, Yinglong; Zheng, Guoyao; Liu, Dequan

    2017-09-01

    The heat flux of the HL-2M divertor would reach 10 MW m-2 or more at the local area when the device operates at high parameters. Subcooled boiling could occur at high thermal load, which would be simulated based on the homogeneous equilibrium model. The results show that the current design of the HL-2M divertor could withstand the local heat flux 10 MW m-2 at a plasma pulse duration of 5 s, inlet coolant pressure of 1.5 MPa and flow velocity of 4 m s-1. The pulse duration that the HL-2M divertor could withstand is closely related to the coolant velocity. In addition, at the time of 2 min after plasma discharge, the flow velocity decreased from 4 m s-1 to 1 m s-1, and the divertor could also be cooled to the initial temperature before the next plasma discharge commences.

  12. On the hydraulic behaviour of ITER Shield Blocks #14 and #08. Computational analysis and comparison with experimental tests

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128, Palermo (Italy); Merola, M.; Mitteau, R.; Raffray, R. [ITER Organization, Route de Vinon sur Verdon, 13067 Saint Paul, Lez Durance (France); Vallone, E., E-mail: eug.vallone@gmail.com [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128, Palermo (Italy)

    2016-11-01

    Highlights: • A benchmarking activity has been carried out focusing the attention on the cooling circuits of ITER Shield Blocks #08 and #14. • A theoretical-computational fluid-dynamic approach based on the Finite Volume Method has been followed, adopting a commercial code. • Hydraulic characteristic functions and spatial distributions of coolant mass flow rate, velocity and pressure drop have been assessed. • Results obtained have allowed code benchmarking for Blanket modules and the numerical predictions have been found to be generally lower than but quite close to the experimental results (lower than 10%). - Abstract: As a consequence of its position and functions, the ITER blanket system will be subjected to significant heat loads under nominal reference conditions. Therefore, the design of its cooling system is particularly demanding. Coolant water is distributed individually to the 440 blanket modules (BMs) through manifold piping, which makes it a highly parallelized system. The mass flow rate distribution is finely tuned to meet all operation constraints: adequate margin to burn out in the plasma facing components, even distribution of water flow among the so-called plasma-facing “fingers” of the Blanket First Wall panels, high enough water flow rate to avoid excessive water temperature in the outlet pipes, maximum allowable water velocity lower than 7 m/s in manifold pipes. Furthermore the overall pressure drop and flow rate in each BM shall be within the fixed specified design limit to avoid an unduly unbalance of cooling among the 440 modules. Analyses have to be carried out following a computational fluid-dynamic (CFD) approach based on the finite volume method and adopting a CFD commercial code to assess the thermal-hydraulic behaviour of each single circuit of the ITER blanket cooling system. This paper describes the code benchmarking needed to determine the best method to get reliable and timely results. Since experimental tests are

  13. Handbook of hydraulic fluid technology

    CERN Document Server

    Totten, George E

    2011-01-01

    ""The Handbook of Hydraulic Fluid Technology"" serves as the foremost resource for designing hydraulic systems and for selecting hydraulic fluids used in engineering applications. Featuring new illustrations, data tables, as well as practical examples, this second edition is updated with essential information on the latest hydraulic fluids and testing methods. The detailed text facilitates unparalleled understanding of the total hydraulic system, including important hardware, fluid properties, and hydraulic lubricants. Written by worldwide experts, the book also offers a rigorous overview of h

  14. ELM - A SIMPLE TOOL FOR THERMAL-HYDRAULIC ANALYSIS OF SOLID-CORE NUCLEAR ROCKET FUEL ELEMENTS

    Science.gov (United States)

    Walton, J. T.

    1994-01-01

    ELM is a simple computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in nuclear thermal rockets. Written for the nuclear propulsion project of the Space Exploration Initiative, ELM evaluates the various heat transfer coefficient and friction factor correlations available for turbulent pipe flow with heat addition. In the past, these correlations were found in different reactor analysis codes, but now comparisons are possible within one program. The logic of ELM is based on the one-dimensional conservation of energy in combination with Newton's Law of Cooling to determine the bulk flow temperature and the wall temperature across a control volume. Since the control volume is an incremental length of tube, the corresponding pressure drop is determined by application of the Law of Conservation of Momentum. The size, speed, and accuracy of ELM make it a simple tool for use in fuel element parametric studies. ELM is a machine independent program written in FORTRAN 77. It has been successfully compiled on an IBM PC compatible running MS-DOS using Lahey FORTRAN 77, a DEC VAX series computer running VMS, and a Sun4 series computer running SunOS UNIX. ELM requires 565K of RAM under SunOS 4.1, 360K of RAM under VMS 5.4, and 406K of RAM under MS-DOS. Because this program is machine independent, no executable is provided on the distribution media. The standard distribution medium for ELM is one 5.25 inch 360K MS-DOS format diskette. ELM was developed in 1991. DEC, VAX, and VMS are trademarks of Digital Equipment Corporation. Sun4 and SunOS are trademarks of Sun Microsystems, Inc. IBM PC is a registered trademark of International Business Machines. MS-DOS is a registered trademark of Microsoft Corporation.

  15. The Numerical Nuclear Reactor for High-Fidelity Integrated Simulation of Neutronic, Thermal-Hydraulic, and Thermo-Mechanical Phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K. S.; Ju, H. G.; Jeon, T. H. and others

    2005-03-15

    A comprehensive high fidelity reactor core modeling capability has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. High fidelity was accomplished by integrating highly refined solution modules for the coupled neutronic, thermal-hydraulic, and thermo-mechanical phenomena. Each solution module employs methods and models that are formulated faithfully to the first-principles governing the physics, real geometry, and constituents. Specifically, the critical analysis elements that are incorporated in the coupled code capability are whole-core neutron transport solution, ultra-fine-mesh computational fluid dynamics/heat transfer solution, and finite-element-based thermo-mechanics solution, all obtained with explicit (fuel pin cell level) heterogeneous representations of the components of the core. The vast computational problem resulting from such highly refined modeling is solved on massively parallel computers, and serves as the 'numerical nuclear reactor'. Relaxation of modeling parameters were also pursued to make problems run on clusters of workstations and PCs for smaller scale applications as well.

  16. Thermal-hydraulic study of the LBE-cooled fuel assembly in the MYRRHA reactor: Experiments and simulations

    Energy Technology Data Exchange (ETDEWEB)

    Pacio, J., E-mail: Julio.pacio@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Wetzel, T. [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Doolaard, H.; Roelofs, F. [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 ZG Petten (Netherlands); Van Tichelen, K. [Belgian Nuclear Reseach Center (SCK-CEN), Boeretang 200, Mol (Belgium)

    2017-02-15

    Heavy liquid metals (HLMs), such as lead-bismuth eutectic (LBE) and pure lead are prominent candidate coolants for many advanced systems based on fast neutrons. In particular, LBE is used in the first-of-its-kind MYRRHA fast reactor, to be built in Mol (Belgium), which can be operated either in critical mode or as a sub-critical accelerator-driven system. With a strong focus on safety, key thermal-hydraulic aspects of these systems, such as the proper cooling of fuel assemblies, must be assessed. Considering the complex geometry and low Prandtl number of LBE (Pr ∼ 0.025), this flow scenario is challenging for the models used in Computational Fluid Dynamics (CFD), e.g. for relating the turbulent transport of momentum and heat. Thus, reliable experimental data for the relevant scenario are needed for validation. In this general context, this topic is studied both experimentally and numerically in the framework of the European FP7 project SEARCH (2011–2015). An experimental campaign, including a 19-rod bundle with wire spacers, cooled by LBE is undertaken at KIT. With prototypical geometry and operating conditions, it is intended to evaluate the validity of current empirical correlations for the MYRRHA conditions and, at the same time, to provide validation data for the CFD simulations performed at NRG. The results of one benchmarking case are presented in this work. Moreover, this validated approach is then used for simulating a complete MYRRHA fuel assembly (127 rods).

  17. The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2014-01-01

    Full Text Available A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised. Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.

  18. Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL

    Directory of Open Access Journals (Sweden)

    A. Rais

    2015-01-01

    Full Text Available In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.

  19. Atucha II NPP full scope simulator modelling with the thermal hydraulic code TRAC{sub R}T

    Energy Technology Data Exchange (ETDEWEB)

    Alonso, Pablo Rey; Ruiz, Jose Antonio; Rivero, Norberto, E-mail: prey@tecnatom.e, E-mail: jaruiz@tecnatom.e, E-mail: nrivero@tecnatom.e [Tecnatom S.A., Madrid (Spain)

    2011-07-01

    In February 2010 NA-SA (Nucleoelectrica Argentina S.A.) awarded Tecnatom the Atucha II full scope simulator project. NA-SA is a public company owner of the Argentinean nuclear power plants. Atucha II is due to enter in operation shortly. Atucha II NPP is a PHWR type plant cooled by the water of the Parana River and has the same design as the Atucha I unit, doubling its power capacity. Atucha II will produce 745 MWe utilizing heavy water as coolant and moderator, and natural uranium as fuel. A plant singular feature is the permanent core refueling. TRAC{sub R}T is the first real time thermal hydraulic six-equations code used in the training simulation industry for NSSS modeling. It is the result from adapting to real time the best estimate code TRACG. TRAC{sub R}T is based on first principle conservation equations for mass, energy and momentum for liquid and steam phases, with two phase flows under non homogeneous and non equilibrium conditions. At present, it has been successfully implemented in twelve full scope replica simulators in different training centers throughout the world. To ease the modeling task, TRAC{sub R}T includes a graphical pre-processing tool designed to optimize this process and alleviate the burden of entering alpha numerical data in an input file. (author)

  20. Implicit Treatment of Technical Specification and Thermal Hydraulic Parameter Uncertainties in Gaussian Process Model to Estimate Safety Margin

    Directory of Open Access Journals (Sweden)

    Douglas A. Fynan

    2016-06-01

    Full Text Available The Gaussian process model (GPM is a flexible surrogate model that can be used for nonparametric regression for multivariate problems. A unique feature of the GPM is that a prediction variance is automatically provided with the regression function. In this paper, we estimate the safety margin of a nuclear power plant by performing regression on the output of best-estimate simulations of a large-break loss-of-coolant accident with sampling of safety system configuration, sequence timing, technical specifications, and thermal hydraulic parameter uncertainties. The key aspect of our approach is that the GPM regression is only performed on the dominant input variables, the safety injection flow rate and the delay time for AC powered pumps to start representing sequence timing uncertainty, providing a predictive model for the peak clad temperature during a reflood phase. Other uncertainties are interpreted as contributors to the measurement noise of the code output and are implicitly treated in the GPM in the noise variance term, providing local uncertainty bounds for the peak clad temperature. We discuss the applicability of the foregoing method to reduce the use of conservative assumptions in best estimate plus uncertainty (BEPU and Level 1 probabilistic safety assessment (PSA success criteria definitions while dealing with a large number of uncertainties.

  1. Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

    Directory of Open Access Journals (Sweden)

    Itamar Iliuk

    2016-01-01

    Full Text Available 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 proposed. A partial loss of flow accident was simulated to show the behavior of the model under this type of accident. The results show that the critical heat flux was detected in the central region along the axial direction of the plate when the right water channel was blocked.

  2. Thermal hydraulics characterization of the core and the reactor vessel type BWR; Caracterizacion termohidraulica del nucleo y de la vasija de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Zapata Y, M.; Lopez H, L.E. [CFE, Carretera Cardel-Nautla Km. 42.5, Municipio Alto Lucero, Veracruz (Mexico)]. e-mail: marxlenin.zapata@cfe.gob.mx

    2008-07-01

    The thermal hydraulics design of a reactor type BWR 5 as the employees in the nuclear power plant of Laguna Verde involves the coupling of at least six control volumes: Pumps jet region, Stratification region, Core region, Vapor dryer region, Humidity separator region and Reactor region. Except by the regions of the core and reactor, these control volumes only are used for design considerations and their importance as operative data source is limited. It is for that is fundamental to complement the thermal hydraulics relations to obtain major data that allow to determine the efficiency of internal components, such as pumps jet, humidity separator and vapor dryer. Like example of the previous thing, calculations are realized on the humidity of the principal vapor during starting, comparing it with the values at the moment incorporated in the data banks of the computers of process of both units. (Author)

  3. 40 CFR 91.329 - Catalyst thermal stress test.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 20 2010-07-01 2010-07-01 false Catalyst thermal stress test. 91.329....329 Catalyst thermal stress test. (a) Oven characteristics. The oven used for termally stressing the test catalyst must be capable of maintaining a temperature of 500 ±5 °C and 1000 ±10 °C. (b) Evaluation...

  4. Modeling the Hydraulic Fracture Stimulation performed for Reservoir Permeability Enhancement at the Grimsel Test Site, Switzerland

    Science.gov (United States)

    Vogler, D.; Settgast, R.; Gischig, V.; Jalali, M.; Doetsch, J.; Valley, B.; Evans, K. F.; Sherman, C.; Saar, M. O.; Amann, F.

    2016-12-01

    In-situ hydraulic stimulation has been performed on the decameter scale in the Deep Underground rock Laboratory (DUG Lab) at the Grimsel Test Site (GTS), Switzerland. The test site consists of granodiorite with a low fracture density and has been well characterized. The GTS is chosen as it represents physical properties representative for crystalline basement where the development of deep enhanced geothermal systems are planned for the future. Conducted stimulation was performed in a number of boreholes, with 3-4 packer intervals in each borehole subjected to repeated stimulation. During each stimulation event, fluid injection pressure, injection flow rate and microseismic events were recorded amongst others. Fully coupled 3D simulations have been performed with the LLNL's GEOS simulation framework. The methods applied in the simulation of the experiments address physical processes such as rock deformation/stress, LEFM fracture mechanics, fluid flow in the fracture and matrix, and the generation of micro-seismic events. This allows investigation in which we may estimate the distance of fracture penetration during the injection phase and correlate the simulated injection pressure with experimental data during injection, as well as post shut-in. Additionally, the extent of the fracture resulting from the numerical model are compared with the spatial distribution of the microseismic events recorded in the experiment.

  5. Hydraulic characterization of overpressured tuffs in central Yucca Flat, Nevada Test Site, Nye County, Nevada

    Science.gov (United States)

    Halford, Keith J.; Laczniak, Randell J.; Galloway, Devin L.

    2005-01-01

    A sequence of buried, bedded, air-fall tuffs has been used extensively as a host medium for underground nuclear tests detonated in the central part of Yucca Flat at the Nevada Test Site. Water levels within these bedded tuffs have been elevated hundreds of meters in areas where underground nuclear tests were detonated below the water table. Changes in the ground-water levels within these tuffs and changes in the rate and distribution of land-surface subsidence above these tuffs indicate that pore-fluid pressures have been slowly depressurizing since the cessation of nuclear testing in 1992. Declines in ground-water levels concurrent with regional land subsidence are explained by poroelastic deformation accompanying ground-water flow as fluids pressurized by underground nuclear detonations drain from the host tuffs into the overlying water table and underlying regional carbonate aquifer. A hydraulic conductivity of about 3 x 10-6 m/d and a specific storage of 9 x 10-6 m-1 are estimated using ground-water flow models. Cross-sectional and three-dimensional ground-water flow models were calibrated to measured water levels and to land-subsidence rates measured using Interferometric Synthetic Aperture Radar. Model results are consistent and indicate that about 2 million m3 of ground water flowed from the tuffs to the carbonate rock as a result of pressurization caused by underground nuclear testing. The annual rate of inflow into the carbonate rock averaged about 0.008 m/yr between 1962 and 2005, and declined from 0.005 m/yr in 2005 to 0.0005 m/yr by 2300.

  6. An approach to modeling coupled thermal-hydraulic-chemical processes in geothermal systems

    Science.gov (United States)

    Palguta, Jennifer; Williams, Colin F.; Ingebritsen, Steven E.; Hickman, Stephen H.; Sonnenthal, Eric

    2011-01-01

    Interactions between hydrothermal fluids and rock alter mineralogy, leading to the formation of secondary minerals and potentially significant physical and chemical property changes. Reactive transport simulations are essential for evaluating the coupled processes controlling the geochemical, thermal and hydrological evolution of geothermal systems. The objective of this preliminary investigation is to successfully replicate observations from a series of hydrothermal laboratory experiments [Morrow et al., 2001] using the code TOUGHREACT. The laboratory experiments carried out by Morrow et al. [2001] measure permeability reduction in fractured and intact Westerly granite due to high-temperature fluid flow through core samples. Initial permeability and temperature values used in our simulations reflect these experimental conditions and range from 6.13 × 10−20 to 1.5 × 10−17 m2 and 150 to 300 °C, respectively. The primary mineralogy of the model rock is plagioclase (40 vol.%), K-feldspar (20 vol.%), quartz (30 vol.%), and biotite (10 vol.%). The simulations are constrained by the requirement that permeability, relative mineral abundances, and fluid chemistry agree with experimental observations. In the models, the granite core samples are represented as one-dimensional reaction domains. We find that the mineral abundances, solute concentrations, and permeability evolutions predicted by the models are consistent with those observed in the experiments carried out by Morrow et al. [2001] only if the mineral reactive surface areas decrease with increasing clay mineral abundance. This modeling approach suggests the importance of explicitly incorporating changing mineral surface areas into reactive transport models.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

  8. Comparison of best-estimate plus uncertainty and conservative methodologies for a PWR MSLB analysis using a coupled 3-D neutron-kinetics/thermal-hydraulic code

    OpenAIRE

    Pericas Casals, Raimon; Ivanov, K.; Reventós Puigjaner, Francesc; Batet Miracle, Lluís

    2017-01-01

    This paper compares the Best-Estimate Plus Uncertainty (BEPU) methodology with the Conservative Bounding methodology for design-basis-accident analysis. Calculations have been performed with TRACE [for thermal-hydraulic (TH) system calculations] and PARCS [for neutron-kinetics (NK) modeling] under the SNAP platform. DAKOTA is used under the SNAP interface for uncertainty and sensitivity analysis. A simplified three-dimensional (3-D) neutronics model of the Ascó II nuclear power plant is used ...

  9. Development and application of the coupled thermal-hydraulics and neutron-kinetics code ATHLET/BIPR-VVER for safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lizorkin, M.; Nikonov, S. [Kurchatov Institute for Atomic Energy, Moscow (Russian Federation); Langenbuch, S.; Velkov, K. [Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany)

    2006-07-01

    The coupled thermal-hydraulics and neutron-kinetics code ATHLET/BIPR-VVER was developed within a co-operation between the RRC Kurchatov Institute (KI) and GRS. The modeling capability of this coupled code as well as the status of validation by benchmark activities and comparison with plant measurements are described. The paper is focused on the modeling of flow mixing in the reactor pressure vessel including its validation and the application for the safety justification of VVER plants. (authors)

  10. The thermal-hydraulic for the new technologies: the micro-fluidics; La thermohydraulique au service des nouvelles technologies: la microfluidique

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

    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.)

  11. Power Control and Monitoring Requirements for Thermal Vacuum/Thermal Balance Testing of the MAP Observatory

    Science.gov (United States)

    Johnson, Chris; Hinkle, R. Kenneth (Technical Monitor)

    2002-01-01

    The specific heater control requirements for the thermal vacuum and thermal balance testing of the Microwave Anisotropy Probe (MAP) Observatory at the Goddard Space Flight Center (GSFC) in Greenbelt, Maryland are described. The testing was conducted in the 10m wide x 18.3m high Space Environment Simulator (SES) Thermal Vacuum Facility. The MAP thermal testing required accurate quantification of spacecraft and fixture power levels while minimizing heater electrical emissions. The special requirements of the MAP test necessitated construction of five (5) new heater racks.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-07-01

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

  13. Combining periodic hydraulic tests and surface tilt measurements to explore in situ fracture hydromechanics

    Science.gov (United States)

    Schuite, Jonathan; Longuevergne, Laurent; Bour, Olivier; Guihéneuf, Nicolas; Becker, Matthew W.; Cole, Matthew; Burbey, Thomas J.; Lavenant, Nicolas; Boudin, Frédéric

    2017-08-01

    Fractured bedrock reservoirs are of socio-economical importance, as they may be used for storage or retrieval of fluids and energy. In particular, the hydromechanical behavior of fractures needs to be understood as it has implications on flow and governs stability issues (e.g., microseismicity). Laboratory, numerical, or field experiments have brought considerable insights to this topic. Nevertheless, in situ hydromechanical experiments are relatively uncommon, mainly because of technical and instrumental limitations. Here we present the early stage development and validation of a novel approach aiming at capturing the integrated hydromechanical behavior of natural fractures. It combines the use of surface tiltmeters to monitor the deformation associated with the periodic pressurization of fractures at depth in crystalline rocks. Periodic injection and withdrawal advantageously avoids mobilizing or extracting significant amounts of fluid, and it hinders any risk of reservoir failure. The oscillatory perturbation is intended to (1) facilitate the recognition of its signature in tilt measurements and (2) vary the hydraulic penetration depth in order to sample different volumes of the fractured bedrock around the inlet and thereby assess scale effects typical of fractured systems. By stacking tilt signals, we managed to recover small tilt amplitudes associated with pressure-derived fracture deformation. Therewith, we distinguish differences in mechanical properties between the three tested fractures, but we show that tilt amplitudes are weakly dependent on pressure penetration depth. Using an elastic model, we obtain fracture stiffness estimates that are consistent with published data. Our results should encourage further improvement of the method.

  14. Development, field testing and implementation of automated hydraulically controlled, variable volume loading systems for reciprocating compressors

    Energy Technology Data Exchange (ETDEWEB)

    Hickman, Dwayne A. [ACI Services, Inc., Cambridge, OH (United States); Slupsky, John [Kvaerner Process Systems, Calgary, Alberta (Canada); Chrisman, Bruce M.; Hurley, Tom J. [Cooper Energy Services, Oklahoma City, OK (United States). Ajax Division

    2003-07-01

    Automated, variable volume unloaders provide the ability to smoothly load/unload reciprocating compressors to maintain ideal operations in ever-changing environments. Potential advantages provided by this load control system include: maximizing unit capacity, optimizing power economy, maintaining low exhaust emissions, and maintaining process suction and discharge pressures. Obstacles foreseen include: reliability, stability, serviceability and automation integration. Results desired include: increased productivity for the compressor and its operators, increased up time, and more stable process control. This presentation covers: system design features with descriptions of how different types of the devices were developed, initial test data, and how they can be effectively operated; three actual-case studies detailing the reasons why automated, hydraulically controlled, variable volume, head-end unloaders were chosen over other types of unloading devices; sophisticated software used in determining the device sizing and predicted performance; mechanical and field considerations; installation, serviceability and operating considerations; device control issues, including PC and PLC considerations; monitoring of actual performance and comparison of such with predicted performance; analysis of mechanical reliability and stability; and preliminary costs versus return on investment analysis. (author)

  15. Hydraulic manipulator design, analysis, and control at Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kress, R.L.; Jansen, J.F. [Oak Ridge National Lab., TN (United States). Robotics and Process Systems Div.; Love, L.J. [Oak Ridge Inst. for Science and Education, TN (United States); Basher, A.M.H. [South Carolina State Univ., Orangeburg, SC (United States)

    1996-09-01

    To meet the increased payload capacities demanded by present-day tasks, manipulator designers have turned to hydraulics as a means of actuation. Hydraulics have always been the actuator of choice when designing heavy-life construction and mining equipment such as bulldozers, backhoes, and tunneling devices. In order to successfully design, build, and deploy a new hydraulic manipulator (or subsystem) sophisticated modeling, analysis, and control experiments are usually needed. To support the development and deployment of new hydraulic manipulators Oak Ridge National Laboratory (ORNL) has outfitted a significant experimental laboratory and has developed the software capability for research into hydraulic manipulators, hydraulic actuators, hydraulic systems, modeling of hydraulic systems, and hydraulic controls. The hydraulics laboratory at ORNL has three different manipulators. First is a 6-Degree-of-Freedom (6-DoF), multi-planer, teleoperated, flexible controls test bed used for the development of waste tank clean-up manipulator controls, thermal studies, system characterization, and manipulator tracking. Finally, is a human amplifier test bed used for the development of an entire new class of teleoperated systems. To compliment the hardware in the hydraulics laboratory, ORNL has developed a hydraulics simulation capability including a custom package to model the hydraulic systems and manipulators for performance studies and control development. This paper outlines the history of hydraulic manipulator developments at ORNL, describes the hydraulics laboratory, discusses the use of the equipment within the laboratory, and presents some of the initial results from experiments and modeling associated with these hydraulic manipulators. Included are some of the results from the development of the human amplifier/de-amplifier concepts, the characterization of the thermal sensitivity of hydraulic systems, and end-point tracking accuracy studies. Experimental and analytical

  16. Thermal analysis of a hypersonic wing test structure

    Science.gov (United States)

    Sandlin, Doral R.; Swanson, Neil J., Jr.

    1989-01-01

    The three-dimensional finite element modeling techniques developed for the thermal analysis of a hypersonic wing test structure (HWTS) are described. The computed results are compared to measured test data. In addition, the results of a NASA two-dimensional parameter finite difference local thermal model and the results of a contractor two-dimensional lumped parameter finite difference local thermal model will be presented.

  17. Hydraulic Properties of a Strike-slip Fault Estimated from Periodic Pumping Tests in Borehole GDP-1 at Grimselpass, Switzerland

    Science.gov (United States)

    Cheng, Y.; Renner, J.

    2016-12-01

    Periodic pumping tests were conducted using a double-packer probe placed at four different depth levels in borehole GDP-1 at Grimselpass, Central Swiss Alps, penetrating a hydrothermally active fault. The tests had the objective to assess the use of periodic testing in the context for design and quality control of hydraulically stimulated geothermal reservoirs. We employed periods between 60 and 1080 s and nominal flow rates between 0 and 10 l/min. Amplitude ratio and phase shift between the pressure and flow rate in the pumping interval corrected for the storage capacity of the tubing are compared to analytical predictions of conventional 1D and radial flow models but also to those for bilinear flow and radial flow in a system composed of cylindrical shells with distinct hydraulic properties concentrically arranged around the well (injectivity analysis). The order of magnitude of derived transmissivity values agrees with that gained from conventional hydraulic tests. The field campaign revealed several advantages of the periodic testing, e.g., short testing time since a periodic signal can easily be separated from changing background pressure by detrending and Fourier transformation. The mismatch between test results and the considered analytical models indicates a hydraulically complex and heterogeneous subsurface. Attenuation and phase shift between the periodic pressure signals above or below packers and in the interval (interference analysis) are addressed by an exploratory numerical modelling of the non-radial flow situation. Interference analysis of a single well provides data that are not affected by uncertainties in the timing of flow-rate changes. The exploratory modeling shows consistent results and improves diffusivity estimation. Yet, more comprehensive modeling will be required to take full advantage of all the pressure records typically acquired when using a double-packer probe.

  18. Thermal cycling tests of actively cooled beryllium copper joints

    Energy Technology Data Exchange (ETDEWEB)

    Roedig, M.; Duwe, R.; Linke, J.; Schuster, A.; Wiechers, B. [Forschungszentrum Juelich GmbH (Germany)

    1998-01-01

    Screening tests (steady state heating) and thermal fatigue tests with several kinds of beryllium-copper joints have been performed in an electron beam facility. Joining techniques under investigation were brazing with silver containing and silver-free braze materials, hot isostatic pressing (HIP) and diffusion bonding (hot pressing). Best thermal fatigue performance was found for the brazed samples. (author)

  19. Thermal-hydraulic modelling of the SAFARI-1 research reactor using RELAP/SCDAPSIM/MOD3.4

    Energy Technology Data Exchange (ETDEWEB)

    Sekhri, Abdelkrim; Graham, Andy [RRT Radiation and Reactor Theory, South African Nuclear Energy Corporation - NECSA, PO Box 582 Pretoria 0001 (South Africa); D' Arcy, Alan; Oliver, Melissa [SAFARI-1 Research Reactor, South African Nuclear Energy Corporation - NECSA, PO Box 582 Pretoria 0001 (South Africa)

    2008-07-01

    The SAFARI-1 reactor is a tank-in-pool MTR type research reactor operated at a nominal core power of 20 MW. It operates exclusively in the single phase liquid water regime with nominal water and fuel temperatures not exceeding 100 deg. C. RELAP/SCDAPSIM/MOD3.4 is a Best Estimate Code for light water reactors as well as for low pressure transients, as part of the code validation was done against low pressure facilities and research reactor experimental data. The code was used to simulate SAFARI-1 in normal and abnormal operation and validated against the experimental data in the plant and was used extensively in the upgrading of the Safety Analysis Report (SAR) of the reactor. The focus of the following study is the safety analysis of the SAFARI-1 research reactor and describes the thermal hydraulic modelling and analysis approach. Particular emphasis is placed on the modelling detail, the application of the no-boiling rule and predicting the Onset of Nucleate Boiling and Departure from Nucleate Boiling under Loss of Flow conditions. Such an event leads the reactor to switch to a natural convection regime which is an adequate mode to maintain the clad and fuel temperature within the safety margin. It is shown that the RELAP/SCDAPSIM/MOD3.4 model can provide accurate predictions as long as the clad temperature remains below the onset of nucleate boiling temperature and the DNB ratio is greater than 2. The results are very encouraging and the model is shown to be appropriate for the analysis of SAFARI-1 research reactor. (authors)

  20. The axial power distribution validation of the SCWR fuel assembly with coupled neutronics-thermal hydraulics method

    Energy Technology Data Exchange (ETDEWEB)

    Xi, Xi [CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China, Chengdu 610041 (China); Xiao, Zejun, E-mail: fabulous_2012@sina.com [CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China, Chengdu 610041 (China); Yan, Xiao; Li, Yongliang; Huang, Yanping [CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China, Chengdu 610041 (China)

    2013-05-15

    Highlights: ► CFX and MCNP codes are suitable to calculate the axial power profile of the FA. ► The partition method in the calculation will affect the final result. ► The density feedback has little effect on the axial power profile of CSR1000 FA. -- Abstract: SCWR (super critical water reactor) is one of the IV generation nuclear reactors in the world. In a typical SCWR the water enters the reactor from the cold leg with a temperature of 280 °C and then leaves the core with a temperature of 500 °C. Due to the sharp change in temperature, there is a huge density change of the water along the axial direction of the fuel assembly (FA), which will affect the moderating power of the water. So the axial power distribution of the SCWR FA could be different from the traditional PWR FA.In this paper, it is the first time that the thermal hydraulics code CFX and neutronics code MCNP are used to analyze the axial power distribution of the SCWR FA. First, the factors in the coupled method which could affect the result are analyzed such as the initialization value or the partition method especially in the MCNP code. Then the axial power distribution of the Europe HPLWR FA is obtained by the coupled method with the two codes and the result is compared with that obtained by Waata and Reiss. There is a good agreement among the three kinds of results. At last, this method is used to calculate the axial power distribution of the Chinese SCWR (CSR1000) FA. It is found the axial power profile of the CSR1000 FA is not so sensitive to the change of the moderator density.

  1. Constraining groundwater discharge in a large watershed: Integrated isotopic, hydraulic, and thermal data from the Canadian shield

    Science.gov (United States)

    Gleeson, Tom; Novakowski, Kent; Cook, Peter G.; Kyser, T. Kurt

    2009-08-01

    The objective of this study is to evaluate the pattern and rate of groundwater discharge in a large, regulated fractured rock watershed using novel and standard methods that are independent of base flow recession. Understanding the rate and pattern of groundwater discharge to surface water bodies is critical for watershed budgets, as a proxy for recharge rates, and for protecting the ecological integrity of lake and river ecosystems. The Tay River is a low-gradient, warm-water river that flows over exposed and fractured bedrock or a thin veneer of coarse-grained sediments. Natural conservative (δ2H, δ18O, Cl, and specific conductance), radioactive (222Rn), and thermal tracers are integrated with streamflow measurements and a steady state advective model to delimit the discharge locations and quantify the discharge fluxes to lakes, wetlands, creeks, and the Tay River. The groundwater discharge rates to most surface water body types are low, indicating that the groundwater and surface water system may be largely decoupled in this watershed compared to watersheds underlain by porous media. Groundwater discharge is distributed across the watershed rather than localized around lineaments or high-density zones of exposed brittle fractures. The results improve our understanding of the rate, localization, and conceptualization of discharge in a large, fractured rock watershed. Applying hydraulic, isotopic, or chemical hydrograph separation techniques would be difficult because the groundwater discharge "signal" is small compared to the "background" surface water inflows or volumes of the surface water bodies. Although this study focuses on a large watershed underlain by fractured bedrock, the methodology developed is transferable to any large regulated or unregulated watershed. The low groundwater discharge rates have significant implications for the ecology, sustainability, and management of large, crystalline watersheds.

  2. Hydraulic impact end effector final test report. Automation and robotics section, ER/WM-AT Program

    Energy Technology Data Exchange (ETDEWEB)

    Couture, S.

    1994-02-18

    One tool being developed for dislodging and fragmenting the hard salt cake waste in the single-shell nuclear waste tanks at the Hanford Reservation near Richland, Washington, is the hydraulic impact end effector (HIEE). This total operates by discharging 11-in. slugs of water at ultrahigh pressures. The HIEE was designed, built, and initially tested in 1992. Work in 1993 included advanced developments of the HIEE to further investigate its fragmentation abilities and to determine more effective operating procedures. These tests showed that more fragmentation can be achieved by increasing the charge pressure of 40 kpsi to 55 kpsi and by the use of different operating procedures. The size of the material and the impact energy of the water slug fired from the HIEE are believed to be major factors in material fragmentation. The material`s ability to fracture also appears to depend on the distance a fracture or crack line must travel to a free surface. Thus, larger material is more difficult to fracture than smaller material. Discharge pressures of 40 kpsi resulted in little penetration or fracturing of the material. At 55 kpsi, however, the size and depth of the fractures increased. Nozzle geometry had a significant effect on fragment size and quantity. Fragmentation was about an order of magnitude greater when the HIEE was discharged into drilled holes rather than onto the material surface. Since surface shots tend to create craters, a multi-shot procedure, coupled with an advanced nozzle design, was used to drill (crater) deep holes into large material. With this procedure, a 600-lb block was reduced to smaller pieces without the use of any additional equipment. Through this advanced development program, the HIEE has demonstrated that it can quickly fragment salt cake material into small, easily removable fragments. The HIEE`s material fragmentation ability can be substantially increased through the use of different nozzle geometries and operating procedures.

  3. CFD Simulation of Thermal-Hydraulic Benchmark V1000CT-2 Using ANSYS CFX

    Directory of Open Access Journals (Sweden)

    Thomas Höhne

    2009-01-01

    Full Text Available Plant measured data from VVER-1000 coolant mixing experiments were used within the OECD/NEA and AER coupled code benchmarks for light water reactors to test and validate computational fluid dynamic (CFD codes. The task is to compare the various calculations with measured data, using specified boundary conditions and core power distributions. The experiments, which are provided for CFD validation, include single loop cooling down or heating-up by disturbing the heat transfer in the steam generator through the steam valves at low reactor power and with all main coolant pumps in operation. CFD calculations have been performed using a numerical grid model of 4.7 million tetrahedral elements. The Best Practice Guidelines in using CFD in nuclear reactor safety applications has been used. Different advanced turbulence models were utilized in the numerical simulation. The results show a clear sector formation of the affected loop at the downcomer, lower plenum and core inlet, which corresponds to the measured values. The maximum local values of the relative temperature rise in the calculation are in the same range of the experiment. Due to this result, it is now possible to improve the mixing models which are usually used in system codes.

  4. Test study of the optimal design for hydraulic performance and treatment performance of free water surface flow constructed wetland.

    Science.gov (United States)

    Guo, Changqiang; Cui, Yuanlai; Dong, Bin; Luo, Yufeng; Liu, Fangping; Zhao, Shujun; Wu, Huirong

    2017-08-01

    Orthogonal tests with mixed levels of design parameters of a free water surface flow constructed wetland were performed to assess their effect on hydraulic and treatment performance, and discover the relationship between the design parameters and the two performances. The results showed that water depth, plant spacing, and layout of in- and outlet mainly affected the two performances. Under 40cm depth, central pass of in- and outlet, 1.8m(3)/h flow rate, 20cm plant spacing, 2:1 aspect ratio, and Scripus tabernaemontani as the plant species, treatment performance of 5.3% TN, 6.1% TP and 15.6% TSS removal efficiencies and a high hydraulic performance of 0.854e, 0.602MI were achieved. There was no significant correlation between the design parameters and the two performances. The relationship among various hydraulic indicators and that among the purification indicators displayed extremely significant correlation. However, there was no significant correlation between hydraulic and treatment performance. Copyright © 2017 Elsevier Ltd. All rights reserved.

  5. IT-Tools Concept for Simulation and Design of Water Hydraulic Mechatronic Test Facilities for Motion Control and Operation in Environmentally Sensitive Application Areas

    DEFF Research Database (Denmark)

    Conrad, Finn; Pobedza, J.; Sobczyk, A.

    2004-01-01

    This paper presents a proposed IT-Tools concept for modeling, simulation, analysis and design of water hydraulic actuators for motion control of machines, lifts, cranes and robots. The designed test rigs have tap water hydraulic components of the Danfoss Nessie® product family and equipped...... with a measurement and data acquisition system. Results of the mathematical modeling, simulation and design of the motion control test rigs are presented. Furthermore, the paper presents selected experimental and identifying test results for the water hydraulic test rigs....

  6. A New Parameter to Assess Hydromechanical Effect in Single-hole Hydraulic Testing and Grouting

    Energy Technology Data Exchange (ETDEWEB)

    Rutqvist, Jonny; Fransson, A.; Tsang, C.-F.; Rutqvist, J.; Gustafson, G.

    2007-09-01

    Grouting or filling of the open voids in fractured rock is done by introducing a fluid, a grout, through boreholes under pressure. The grout may be either a Newtonian fluid or a Bingham fluid. The penetration of the grout and the resulting pressure profile may give rise to hydromechanical effects, which depends on factors such as the fracture aperture, pressure at the borehole and the rheological properties of the grout. In this paper, we postulate that a new parameter, {angstrom}, which is the integral of the fluid pressure change in the fracture plane, is an appropriate measure to describe the change in fracture aperture volume due to a change in effective stress. In many cases, analytic expressions are available to calculate pressure profiles for relevant input data and the {angstrom} parameter. The approach is verified against a fully coupled hydromechanical simulator for the case of a Newtonian fluid. Results of the verification exercise show that the new approach is reasonable and that the {angstrom}-parameter is a good measure for the fracture volume change: i.e., the larger the {angstrom}-parameter, the larger the fracture volume change, in an almost linear fashion. To demonstrate the application of the approach, short duration hydraulic tests and constant pressure grouting are studied. Concluded is that using analytic expressions for penetration lengths and pressure profiles to calculate the {angstrom} parameter provides a possibility to describe a complex situation and compare, discuss and weigh the impact of hydromechanical couplings for different alternatives. Further, the analyses identify an effect of high-pressure grouting, where uncontrolled grouting of larger fractures and insufficient (or less-than-expected) sealing of finer fractures is a potential result.

  7. Contribution to the study of thermal-hydraulic problems in nuclear reactors; Contribution a l`etude de problemes de thermohydraulique dans les reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Cognet, G

    1998-07-07

    In nuclear reactors, whatever the type considered, Pressurized Water Water Reactors (PWRs), Fast Breeder reactors (FBRs)..., thermal-hydraulics, the science of fluid mechanics and thermal behaviour, plays an essential role, both in nominal operating and accidental conditions. Fluid can either be the primary fluid (liquid or gas) or a very specific fluid called corium, which, in case of severe accident, could result from core and environning structure melting. The work reported here represents a 20-year contribution to thermal-hydraulic issues which could occur in FBRs and PWRs. Working on these two types of reactors, both in nominal and severe accident situations, has allowed me to compare the problems and to realize the importance of communication between research teams. The evolution in the complexity of studied problems, unavoidable in order to reduce costs and significantly improve safety, has led me from numerical modelling of single-phase flow turbulence to high temperature real melt experiments. The difficulties encountered in understanding the observed phenomena and in increasing experimental databases for computer code qualification have often entailed my participation in specific measurement device developments or adaptations, in particular non-intrusive devices generally based on optical techniques. Being concerned about the end-use of this research work, I actively participated in `in-situ` thermalhydraulic experiments in the FBRs: Phenix and Super-Phenix, of which I appreciated their undeniable scientific contribution. In my opinion, the thermal-hydraulic questions related to severe accidents are the most complex as they are at the cross-roads of several scientific specialities. Consequently, they require a multi-disciplinary approach and a continuous see-saw motion between experimentalists and modelling teams. After a brief description of the various problems encountered, the main ones are reported. Finally, the importance for research teams to

  8. Thermal and Alignment Analysis of the Instrument-Level ATLAS Thermal Vacuum Test

    Science.gov (United States)

    Bradshaw, Heather

    2012-01-01

    This paper describes the thermal analysis and test design performed in preparation for the ATLAS thermal vacuum test. NASA's Advanced Topographic Laser Altimeter System (ATLAS) will be flown as the sole instrument aboard the Ice, Cloud, and land Elevation Satellite-2 (ICESat-2). It will be used to take measurements of topography and ice thickness for Arctic and Antarctic regions, providing crucial data used to predict future changes in worldwide sea levels. Due to the precise measurements ATLAS is taking, the laser altimeter has very tight pointing requirements. Therefore, the instrument is very sensitive to temperature-induced thermal distortions. For this reason, it is necessary to perform a Structural, Thermal, Optical Performance (STOP) analysis not only for flight, but also to ensure performance requirements can be operationally met during instrument-level thermal vacuum testing. This paper describes the thermal model created for the chamber setup, which was used to generate inputs for the environmental STOP analysis. This paper also presents the results of the STOP analysis, which indicate that the test predictions adequately replicate the thermal distortions predicted for flight. This is a new application of an existing process, as STOP analyses are generally performed to predict flight behavior only. Another novel aspect of this test is that it presents the opportunity to verify pointing results of a STOP model, which is not generally done. It is possible in this case, however, because the actual pointing will be measured using flight hardware during thermal vacuum testing and can be compared to STOP predictions.

  9. Irradiation and testing of off-the-shelf seal materials for water hydraulic applications in ITER remote handling equipment

    Energy Technology Data Exchange (ETDEWEB)

    Irving, M. [EFDA-CSU Garching, Boltzmannstrasse 2, 85748 Garching (Germany)], E-mail: mike.irving@tech.efda.org; Tammisto, J. [Tampere University of Technology, PO Box 589, 33101 Tampere (Finland); Hodgson, E.R.; Hernandez, T. [Euratom/CIEMAT Fusion Association, Avenida Complutense 22, 28040 Madrid (Spain)

    2007-10-15

    Remote handling (RH) is one of the most challenging aspects of the ITER project, and the European home team is building a major prototype of the divertor region (the Divertor Test Platform 2) to confirm practically the RH concepts proposed in this area. To handle the 9 Tonne divertor cassette, water hydraulics has been selected because it offers high forces and precise control in a compact envelope, with minimal long-term contamination should a leak develop. Water hydraulic components use mainly stainless steel - unaffected by gamma radiation - but the integral seals and O-rings are known to be sensitive. For radiation testing of these components, a modular approach was adopted, enabling up to 11 seal carriers assemblies to be irradiated simultaneously in the limited space available, with individual carriers being removed at varying total doses up to 10 MGy. Each carrier was then installed in a real hydraulic rig for testing, revealing not only at what total dose the components became unusable, but also how they fail, enabling condition monitoring to assess the state of the seals long before their failure might render the RH equipment irrecoverable.

  10. Multi-level slug tests in highly permeable formations: 2. Hydraulic conductivity identification, method verification, and field applications

    Science.gov (United States)

    Zlotnik, V.A.; McGuire, V.L.

    1998-01-01

    Using the developed theory and modified Springer-Gelhar (SG) model, an identification method is proposed for estimating hydraulic conductivity from multi-level slug tests. The computerized algorithm calculates hydraulic conductivity from both monotonic and oscillatory well responses obtained using a double-packer system. Field verification of the method was performed at a specially designed fully penetrating well of 0.1-m diameter with a 10-m screen in a sand and gravel alluvial aquifer (MSEA site, Shelton, Nebraska). During well installation, disturbed core samples were collected every 0.6 m using a split-spoon sampler. Vertical profiles of hydraulic conductivity were produced on the basis of grain-size analysis of the disturbed core samples. These results closely correlate with the vertical profile of horizontal hydraulic conductivity obtained by interpreting multi-level slug test responses using the modified SG model. The identification method was applied to interpret the response from 474 slug tests in 156 locations at the MSEA site. More than 60% of responses were oscillatory. The method produced a good match to experimental data for both oscillatory and monotonic responses using an automated curve matching procedure. The proposed method allowed us to drastically increase the efficiency of each well used for aquifer characterization and to process massive arrays of field data. Recommendations generalizing this experience to massive application of the proposed method are developed.Using the developed theory and modified Springer-Gelhar (SG) model, an identification method is proposed for estimating hydraulic conductivity from multi-level slug tests. The computerized algorithm calculates hydraulic conductivity from both monotonic and oscillatory well responses obtained using a double-packer system. Field verification of the method was performed at a specially designed fully penetrating well of 0.1-m diameter with a 10-m screen in a sand and gravel alluvial

  11. Hydraulic Yaw System

    DEFF Research Database (Denmark)

    Stubkier, Søren; Pedersen, Henrik C.; Mørkholt, M.

    system and rotor shaft when utilizing the soft yaw drive concept compared to the original stiff yaw system. The physical demands of the hydraulic yaw system are furthermore examined for a life time of 20 years. Based on the extrapolated loads, the duty cycles show that it is possible to construct...... a hydraulic soft yaw system, which is able to reduce the loads on the wind turbine significantly. A full scale hydraulic yaw test rig is available for experiments and tests. The test rig is presented as well as the system schematics of the hydraulic yaw system....

  12. A new coupling of the 3D thermal-hydraulic code THYC and the thermo-mechanical code CYRANO3 for PWR calculations

    Energy Technology Data Exchange (ETDEWEB)

    Marguet, S.D. [Electricite de France (EDF), 92 - Clamart (France)

    1997-12-31

    Among all parameters, the fuel temperature has a significant influence on the reactivity of the core, because of the Doppler effect on cross-sections. Most neutronic codes use a straightforward method to calculate an average fuel temperature used in their specific feed-back models. For instance, EDF`s neutronic code COCCINELLE uses the Rowland`s formula using the temperatures of the center and the surface of the pellet. COCCINELLE is coupled to the 3D thermal-hydraulic code THYC with calculates TDoppler with is standard thermal model. In order to improve the accuracy of such calculations, we have developed the coupling of our two latest codes in thermal-hydraulics (THYC) and thermo-mechanics (CYRANO3). THYC calculates two-phase flows in pipes or rod bundles and is used for transient calculations such as steam-line break, boron dilution accidents, DNB predictions, steam generator and condenser studies. CYRANO3 calculates most of the phenomena that take place in the fuel such as: 1) heat transfer induced by nuclear power; 2) thermal expansion of the fuel and the cladding; 3) release of gaseous fission`s products; 4) mechanical interaction between the pellet and the cladding. These two codes are now qualified in their own field and the coupling, using Parallel Virtual Machine (PVM) libraries customized in an home-made-easy-to-use package called CALCIUM, has been validated on `low` configurations (no thermal expansion, constant thermal characteristics) and used on accidental transients such as rod ejection and loss of coolant accident. (K.A.) 7 refs.

  13. Data from thermal testing of the Open Source Cryostage

    DEFF Research Database (Denmark)

    Buch, Johannes Lørup; Ramløv, Hans

    2016-01-01

    The data presented here is related to the research article "An open source cryostage and software analysis method for detection of antifreeze activity" (Buch and Ramløv, 2016) [1]. The design of the Open Source Cryostage (OSC) is tested in terms of thermal limits, thermal efficiency and electrical...

  14. Evaluation of Fire Test Methods for Aircraft Thermal Acoustical Insulation

    Science.gov (United States)

    1997-09-01

    This report presents the results of laboratory round robin flammability testing performed on thermal acoustical insulation blankets and the films used as insulation coverings. This work was requested by the aircraft industry as a result of actual inc...

  15. Influence of stone content on soil hydraulic properties: experimental investigation and test of existing model concepts

    Science.gov (United States)

    Naseri, Mahyar; Richter, Niels; Iden, Sascha C.; Durner, Wolfgang

    2017-04-01

    Rock fragments in soil, in this contribution referred to as "stones", play an important role for water flow in the subsurface. To successfully model soil hydraulic processes such as evaporation, redistribution and drainage, an understanding of how stones affect soil hydraulic properties (SHP) is crucial. Past investigations on the role of stones in soil have focused on their influence on the water retention curve (WRC) and on saturated hydraulic conductivity Ks, and have led to some simple theoretical models for the influence of stones on effective SHP. However, studies that measure unsaturated SHP directly, i.e., simultaneously the WRC and hydraulic conductivity curve (HCC) are still missing. Also, studies so far were restricted to low or moderate stone contents of less than 40%. We conducted a laboratory study in which we examined the effect of stone content on effective WRC and HCC of stony soils. Mixtures of soil and stones were generated by substituting background soil with stones in weight fractions between 0% (fine material only) to 100% (pure gravel). Stone sizes were 2-5 mm and 7-15 mm, respectively, and background soils were Sand and Sandy Loam. Packed samples were fully saturated under vacuum and subsequently subjected to evaporation in the laboratory. All experiments were done in three replicates. The soil hydraulic properties were determined by the simplified evaporation method using the UMS HYPROP setup. Questions were whether the applied measurement methodology is applicable to derive the SHP of the mixtures and how the gradual increase of stone content will affect the SHP, particularly the HCC. The applied methodology was successful in identifying effective SHP with a high precision over the full moisture range. WRC and HCC were successfully obtained by HYPROP, even for pure gravel with a size of 7-15 mm. WRCs changed qualitatively in the expected manner, i.e., an increase of stone content reduced porosity and soil water content at all suctions

  16. Rocket nozzle thermal shock tests in an arc heater facility

    Science.gov (United States)

    Painter, James H.; Williamson, Ronald A.

    1986-01-01

    A rocket motor nozzle thermal structural test technique that utilizes arc heated nitrogen to simulate a motor burn was developed. The technique was used to test four heavily instrumented full-scale Star 48 rocket motor 2D carbon/carbon segments at conditions simulating the predicted thermal-structural environment. All four nozzles survived the tests without catastrophic or other structural failures. The test technique demonstrated promise as a low cost, controllable alternative to rocket motor firing. The technique includes the capability of rapid termination in the event of failure, allowing post-test analysis.

  17. Test facility of thermal storage equipment for space power generation

    Science.gov (United States)

    Inoue, T.; Nakagawa, M.; Mochida, Y.; Ohtomo, F.; Shimizu, K.; Tanaka, K.; Abe, Y.; Nomura, O.; Kamimoto, M.

    A thermal storage equipment test facility has been built in connection with developing solar dynamic power systems (SDPSs). The test facility consists of a recuperative closed Brayton cycle system (CBC), with a mixture of helium and xenon with a molecular weight of 39.9 serving as the working fluid. CBC has been shown to be the most attractive power generation system among several types of SDPSs because of its ability to meet the required high power demand and its thermal efficiency, about 30 percent. The authors present a description of this test facility and give results of the preliminary test and the first-stage test with heat storage equipment.

  18. Liquid Metal Thermal Electric Converter bench test module

    Energy Technology Data Exchange (ETDEWEB)

    Lukens, L.L.; Andraka, C.E.; Moreno, J.B.

    1988-04-01

    This report describes the design, fabrication, and test of a Liquid Metal Thermal Electric Converter Bench Test Module. The work presented in this document was conducted as a part of Heat Engine Task of the US Department of Energy's (DOE) Solar Thermal Technology Program. The objective of this task is the development and evaluation of heat engine technologies applicable to distributed receiver systems, in particular, dish electric systems.

  19. Significance of Landsat-7 Spacecraft Level Thermal Balance and Thermal Test for ETM+Instrument

    Science.gov (United States)

    Choi, Michael K.

    1999-01-01

    The thermal design and the instrument thermal vacuum (T/V) test of the Landsat-7 Enhanced Thematic Mapper Plus (ETM+) instrument were based on the Landsat-4, 5 and 6 heritage. The ETM+ scanner thermal model was also inherited from Landsat-4, 5 and 6. The temperature predictions of many scanner components in the original thermal model had poor agreement with the spacecraft and instrument integrated sun-pointing safehold (SPSH) thermal balance (T/B) test results. The spacecraft and instrument integrated T/B test led to a change of the Full Aperture Calibrator (FAC) motor stack "solar shield" coating from MIL-C-5541 to multi-layer insulation (MLI) thermal blanket. The temperature predictions of the Auxiliary Electronics Module (AEM) in the thermal model also had poor agreement with the T/B test results. Modifications to the scanner and AEM thermal models were performed to give good agreement between the temperature predictions and the test results. The correlated ETM+ thermal model was used to obtain flight temperature predictions. The flight temperature predictions in the nominal 15-orbit mission profile, plus margins, were used as the yellow limits for most of the ETM+ components. The spacecraft and instrument integrated T/B and TN test also revealed that the standby heater capacity on the Scan Mirror Assembly (SMA) was insufficient when the Earth Background Simulator (EBS) was 1 50C or colder, and the baffle heater possibly caused the coherent noise in the narrow band data when it was on. Also, the cooler cool-down was significantly faster than that in the instrument T/V test, and the coldest Cold Focal Plane Array (CFPA) temperature achieved was colder.

  20. Groundwater Flow and Thermal Modeling to Support a Preferred Conceptual Model for the Large Hydraulic Gradient North of Yucca Mountain

    Energy Technology Data Exchange (ETDEWEB)

    McGraw, D.; Oberlander, P.

    2007-12-18

    The purpose of this study is to report on the results of a preliminary modeling framework to investigate the causes of the large hydraulic gradient north of Yucca Mountain. This study builds on the Saturated Zone Site-Scale Flow and Transport Model (referenced herein as the Site-scale model (Zyvoloski, 2004a), which is a three-dimensional saturated zone model of the Yucca Mountain area. Groundwater flow was simulated under natural conditions. The model framework and grid design describe the geologic layering and the calibration parameters describe the hydrogeology. The Site-scale model is calibrated to hydraulic heads, fluid temperature, and groundwater flowpaths. One area of interest in the Site-scale model represents the large hydraulic gradient north of Yucca Mountain. Nearby water levels suggest over 200 meters of hydraulic head difference in less than 1,000 meters horizontal distance. Given the geologic conceptual models defined by various hydrogeologic reports (Faunt, 2000, 2001; Zyvoloski, 2004b), no definitive explanation has been found for the cause of the large hydraulic gradient. Luckey et al. (1996) presents several possible explanations for the large hydraulic gradient as provided below: The gradient is simply the result of flow through the upper volcanic confining unit, which is nearly 300 meters thick near the large gradient. The gradient represents a semi-perched system in which flow in the upper and lower aquifers is predominantly horizontal, whereas flow in the upper confining unit would be predominantly vertical. The gradient represents a drain down a buried fault from the volcanic aquifers to the lower Carbonate Aquifer. The gradient represents a spillway in which a fault marks the effective northern limit of the lower volcanic aquifer. The large gradient results from the presence at depth of the Eleana Formation, a part of the Paleozoic upper confining unit, which overlies the lower Carbonate Aquifer in much of the Death Valley region. The

  1. INVESTIGATION OF HYDRAULIC OPERATIONAL REGIMES OF CIRCULATING SYSTEM AT THERMAL POWER PLANT OF VOLZHSKY AUTOMOTIVE WORKS USING COMPUTER MODEL

    Directory of Open Access Journals (Sweden)

    V. V. Dikop

    2005-01-01

    Full Text Available Investigation results of hydraulic operational regimes of a circulating system with the help of a computer model are presented in the paper. The models simulates hydraulic processes by means of an iterated solution of a set of algebraic non-linear equations that is formed while using a graph theory The circulating system is considered as a unit in the model. The model program makes it possible to calculate consumption and pressure at any point of the circulating system with indication of flow motion directions along its separate branches and also to execute some economical calculations.

  2. Hydraulic Property and Soil Textural Classification Measurements for Rainier Mesa, Nevada Test Site, Nevada

    Science.gov (United States)

    Ebel, Brian A.; Nimmo, John R.

    2010-01-01

    This report presents particle size analysis, field-saturated hydraulic conductivity measurements, and qualitative descriptions of surficial materials at selected locations at Rainier Mesa, Nevada. Measurements and sample collection were conducted in the Rainier Mesa area, including unconsolidated sediments on top of the mesa, an ephemeral wash channel near the mesa edge, and dry U12n tunnel pond sediments below the mesa. Particle size analysis used a combination of sieving and optical diffraction techniques. Field-saturated hydraulic conductivity measurements employed a single-ring infiltrometer with analytical formulas that correct for falling head and spreading outside the ring domain. These measurements may prove useful to current and future efforts at Rainier Mesa aimed at understanding infiltration and its effect on water fluxes and radionuclide transport in the unsaturated zone.

  3. Hydraulic Property and Soil Textural Classification Measurements for Rainier Mesa, Nevada Test Site, Nevada

    Energy Technology Data Exchange (ETDEWEB)

    Ebel, Brian A.; Nimmo, John R.

    2009-12-29

    This report presents particle size analysis, field-saturated hydraulic conductivity measurements, and qualitative descriptions of surficial materials at selected locations at Rainier Mesa, Nevada. Measurements and sample collection were conducted in the Rainier Mesa area, including unconsolidated sediments on top of the mesa, an ephemeral wash channel near the mesa edge, and dry U12n tunnel pond sediments below the mesa. Particle size analysis used a combination of sieving and optical diffraction techniques. Field-saturated hydraulic conductivity measurements employed a single-ring infiltrometer with analytical formulas that correct for falling head and spreading outside the ring domain. These measurements may prove useful to current and future efforts at Rainier Mesa aimed at understanding infiltration and its effect on water fluxes and radionuclide transport in the unsaturated zone.

  4. Arc Jet Testing of Thermal Protection Materials: 3 Case Studies

    Science.gov (United States)

    Johnson, Sylvia; Conley, Joe

    2015-01-01

    Arc jet testing is used to simulate entry to test thermal protection materials. This paper discusses the usefulness of arc jet testing for 3 cases. Case 1 is MSL and PICA, Case 2 is Advanced TUFROC, and Case 3 is conformable ablators.

  5. Advanced thermal energy management: A thermal test bed and heat pipe simulation

    Science.gov (United States)

    Barile, Ronald G.

    1986-01-01

    Work initiated on a common-module thermal test simulation was continued, and a second project on heat pipe simulation was begun. The test bed, constructed from surplus Skylab equipment, was modeled and solved for various thermal load and flow conditions. Low thermal load caused the radiator fluid, Coolanol 25, to thicken due to its temperature avoided by using a regenerator-heat-exchanger. Other possible solutions modeled include a radiator heater and shunting heat from the central thermal bus to the radiator. Also, module air temperature can become excessive with high avionics load. A second preoject concerning advanced heat pipe concepts was initiated. A program was written which calculates fluid physical properties, liquid and vapor pressure in the evaporator and condenser, fluid flow rates, and thermal flux. The program is directed to evaluating newer heat pipe wicks and geometries, especially water in an artery surrounded by six vapor channels. Effects of temperature, groove and slot dimensions, and wick properties are reported.

  6. IBL Thermal Mockup Bake-Out Tests

    CERN Document Server

    Nuiry, FX

    2014-01-01

    This note summarizes different bake-out tests that have been performed with the ATLAS Insertable B-Layer (IBL) mockup. Two beam pipe configurations have been tested: one with the aerogel insulation layer all along the pipe and one without insulation over 622 mm around Z0. These tests have been crucial for decisions about aerogel removal, choice of heaters for the LHC beam pipe bake-out, and choice of temperature setpoints for the cooling system during nominal IBL operation. They also revealed very useful information on integration issues and the thermo-mechanical behaviour of the IBL detector.

  7. Thermal Analysis of Bending Under Tension Test

    DEFF Research Database (Denmark)

    Ceron, Ermanno; Martins, Paulo A.F.; Bay, Niels

    2014-01-01

    sometimes can cause lubricant film breakdown and galling. In order to replicate the production conditions in bending under tension testing it is thus important to control the tool/workpiece interface temperature. This can be done by pre-heating the tool, but it is essential that the interface temperature......The tribological conditions in deep drawing can be simulated in the Bending Under Tension test to evaluate the performance of new lubricants, tool materials, etc. Deep drawing production with automatic handling runs normally at high rate. This implies considerable heating of the tools, which...... during testing is similar to the one in the production tool. A universal sheet tribo-tester has been developed, which can run multiple tests automatically from coil. This allows emulating the temperature increase as in production. The present work performs finite element analysis of the evolution...

  8. 5-Megawatt solar-thermal test facility: environmental assessment

    Energy Technology Data Exchange (ETDEWEB)

    None

    1976-01-30

    An Environmental Assessment of the 5 Megawatt Solar Thermal Test Facility (STTF) is presented. The STTF is located at Albuquerque, New Mexico. The facility will have the capability for testing scale models of major subsystems comprising a solar thermal electrical power plant. The STTF capabilities will include testing a solar energy collector subsystem comprised of heliostat arrays, a receiver subsystem which consists of a boiler/superheater in which a working fluid is heated, and a thermal storage subsystem which includes tanks of high heat capacity material which stores thermal energy for subsequent use. The STTF will include a 200-foot receiver tower on which experimental receivers will be mounted. The Environmental Assessment describes the proposed STTF, its anticipated benefits, and the environment affected. It also evaluates the potential environmental impacts associated with STTF construction and operation.

  9. Transient Thermal Testing and Analysis of a Thermally Insulating Structural Sandwich Panel

    Science.gov (United States)

    Blosser, Max L.; Daryabeigi, Kamran; Bird, Richard K.; Knutson, Jeffrey R.

    2015-01-01

    A core configuration was devised for a thermally insulating structural sandwich panel. Two titanium prototype panels were constructed to illustrate the proposed sandwich panel geometry. The core of one of the titanium panels was filled with Saffil(trademark) alumina fibrous insulation and the panel was tested in a series of transient thermal tests. Finite element analysis was used to predict the thermal response of the panel using one- and two-dimensional models. Excellent agreement was obtained between predicted and measured temperature histories.

  10. Thermal Vacuum/Balance Test Results of Swift BAT with Loop Heat Pipe Thermal System

    Science.gov (United States)

    Choi, Michael K.

    2004-01-01

    The Swift Burst Alert Telescope (BAT) Detector Array is thermally well coupled to eight constant conductance heat pipes (CCHPs) embedded in the Detector Array Plate PAP), and two loop heat pipes (LHPs) transport heat from the CCHPs to a radiator. The CCHPs have ammonia as the working fluid and the LHPs have propylene as the working fluid. Precision heater controllers, which have adjustable set points in flight, are used to control the LHP compensation chamber and Detector Array xA1 ASIC temperatures. The radiator has AZ-Tek's AZW-LA-II low solar absorptance white paint as the thermal coating, and is located on the anti-sun side of the spacecraft. A thermal balance (T/B) test on the BAT was successfully completed. It validated that the thermal design satisfies the temperature requirements of the BAT in the flight thermal environments. Instrument level and observatory level thermal vacuum (TN) cycling tests of the BAT Detector Array by using the LHP thermal system were successfully completed. This paper presents the results of the T/B test and T N cycling tests.

  11. Experimental-based Modelling and Simulation of Water Hydraulic Mechatronics Test Facilities for Motion Control and Operation in Environmental Sensitive Applications` Areas

    DEFF Research Database (Denmark)

    Conrad, Finn; Pobedza, J.; Sobczyk, A.

    2003-01-01

    test rig facilities powered by environmental friendly water hydraulic servo actuator system. Test rigs with measurement and data acquisition system were designed and build up with tap water hydraulic components of the Danfoss Nessie® product family. This paper presents selected experimental......The paper presents experimental-based modelling, simulation, analysis and design of water hydraulic actuators for motion control of machines, lifts, cranes and robots. The contributions includes results from on-going research projects on fluid power and mechatronics based on tap water hydraulic...... proportional valves and servo actuators for motion control and power transmission undertaken in co-operation by Technical University, DTU and Cracow University of Technology, CUT. The results of this research co-operation include engineering design and test of simulation models compared with two mechatronic...

  12. The test facility requirements for the thermal vacuum thermal balance test of the Cosmic Background Explorer Observatory

    Science.gov (United States)

    Milam, Laura J.

    1990-01-01

    The Cosmic Background Explorer Observatory (COBE) underwent a thermal vacuum thermal balance test in the Space Environment Simulator (SES). This was the largest and most complex test ever conducted at this facility. The 4 x 4 m (13 x 13 ft) spacecraft weighed approx. 2223 kg (4900 lbs) for the test. The test set up included simulator panels for the inboard solar array panels, simulator panels for the flight cowlings, Sun and Earth Sensor stimuli, Thermal Radio Frequency Shield heater stimuli and a cryopanel for thermal control in the Attitude Control System Shunt Dissipator area. The fixturing also included a unique 4.3 m (14 ft) diameter Gaseous Helium Cryopanel which provided a 20 K environment for the calibration of one of the spacecraft's instruments, the Differential Microwave Radiometer. This cryogenic panel caused extra contamination concerns and a special method was developed and written into the test procedure to prevent the high buildup of condensibles on the panel which could have led to backstreaming of the thermal vacuum chamber. The test was completed with a high quality simulated space environment provided to the spacecraft. The test requirements, test set up, and special fixturing are described.

  13. Uncertainty analysis for results of thermal hydraulic codes of best-estimate-type; Analisis de incertidumbre para resultados de codigos termohidraulicos de mejor estimacion

    Energy Technology Data Exchange (ETDEWEB)

    Alva N, J.

    2010-07-01

    In this thesis, some fundamental knowledge is presented about uncertainty analysis and about diverse methodologies applied in the study of nuclear power plant transient event analysis, particularly related to thermal hydraulics phenomena. These concepts and methodologies mentioned in this work come from a wide bibliographical research in the nuclear power subject. Methodologies for uncertainty analysis have been developed by quite diverse institutions, and they have been widely used worldwide for application to results from best-estimate-type computer codes in nuclear reactor thermal hydraulics and safety analysis. Also, the main uncertainty sources, types of uncertainties, and aspects related to best estimate modeling and methods are introduced. Once the main bases of uncertainty analysis have been set, and some of the known methodologies have been introduced, it is presented in detail the CSAU methodology, which will be applied in the analyses. The main objective of this thesis is to compare the results of an uncertainty and sensibility analysis by using the Response Surface Technique to the application of W ilks formula, apply through a loss coolant experiment and an event of rise in a BWR. Both techniques are options in the part of uncertainty and sensibility analysis of the CSAU methodology, which was developed for the analysis of transients and accidents at nuclear power plants, and it is the base of most of the methodologies used in licensing of nuclear power plants practically everywhere. Finally, the results of applying both techniques are compared and discussed. (Author)

  14. Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi; Zhang, Dingkang

    2013-11-29

    This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implement a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis.

  15. A Thermal Test System for Helmet Cooling Studies

    Directory of Open Access Journals (Sweden)

    Shaun Fitzgerald

    2018-02-01

    Full Text Available One of the primary causes of discomfort to both irregular and elite cyclists is heat entrapment by a helmet resulting in overheating and excessive sweating of the head. To accurately assess the cooling effectiveness of bicycle helmets, a heated plastic thermal headform has been developed. The construction consists of a 3D-printed headform of low thermal conductivity with an internal layer of high thermal mass that is heated to a constant uniform temperature by an electrical heating element. Testing is conducted in a wind tunnel where the heater power remains constant and the resulting surface temperature distribution is directly measured by 36 K-type thermocouples embedded within the surface of the head in conjunction with a thermal imaging camera. Using this new test system, four bicycle helmets were studied in order to measure their cooling abilities and to identify ‘hot spots’ where cooling performance is poor.

  16. Data from thermal testing of the Open Source Cryostage

    DEFF Research Database (Denmark)

    Buch, Johannes Lørup; Ramløv, Hans

    2016-01-01

    The data presented here is related to the research article “An open source cryostage and software analysis method for detection of antifreeze activity” (Buch and Ramløv, 2016) [1]. The design of the Open Source Cryostage (OSC) is tested in terms of thermal limits, thermal efficiency and electrical...... efficiency. This article furthermore includes an overview of the electrical circuitry and a flowchart of the software program controlling the temperature of the OSC. The thermal efficiency data is presented here as degrees per volt and maximum cooling capacity....

  17. COMPARISON OF RESULTS OF THERMAL TESTS OF BALCONY DOORS

    Directory of Open Access Journals (Sweden)

    Golubev Stanislav Sergeevich

    2012-10-01

    Full Text Available Results of thermal tests of balcony doors are presented in the article. In the course of the research project, two types of doors were tested. The first type represents a PVC frame door (width 82 mm; it has a triple glazing (4K-16Ar-4-16Ar-K4; its blank part represents a polystyrene sandwich panel (width 40 mm. The second type represents a PVC frame door (width 82 mm, that has a triple glazing (4K-16Ar-4-16Ar-K4 and composite PVC panels. The testing procedure and processing results are described in the article. The test has demonstrated that the thermal resistance value of the balcony door of the first type exceeds the thermal resistance value of the balcony door of the second type.

  18. Development of a model for the thermal-hydraulic characterization of the He-FUS3 loop

    Energy Technology Data Exchange (ETDEWEB)

    Barone, G., E-mail: gianluca.barone@for.unipi.it [University of Pisa, Department of Civil and Industrial Engineering (DICI), Pisa (Italy); Coscarelli, E.; Forgione, N.; Martelli, D. [University of Pisa, Department of Civil and Industrial Engineering (DICI), Pisa (Italy); Del Nevo, A.; Tarantino, M.; Utili, M. [ENEA UTIS-TCI, CR Brasimone, Camugnano (Italy); Ricapito, I.; Calderoni, P. [TBM& MD Project, Fusion for Energy, EU Commission, Carrer J. Pla, 2, Building B3, 08019 Barcelona (Spain)

    2015-10-15

    Highlights: • RELAP5-3D model of He-FUS3 facility and turbo circulator (TC). • Cold and hot facility T/H numerical analysis for the TC operanting range. • Effect of the cold by-pass opening on the facility performances. - Abstract: He-FUS3 is a helium facility designed and realized by ENEA in order to test the thermal-mechanical properties of prototypical breeding blanket module assemblies of a DEMO reactor. The actual facility has been upgraded with a high performance turbo circulator and a water heat exchanger integrating the pre-existent air cooler. In addition, a new test section located in the loop hot zone has been settled down with the objective of investigating safety relevant transient conditions of “In-TBM” LOCA scenarios. A RELAP5-3D{sup ©} model has been developed to perform a set of preliminary simulations on the new He-FUS3 layout. Both cold and hot stationary conditions have been analyzed evaluating the turbo circulator performances for a wide range of helium flow rate. Outcomes have shown that RELAP5-3D{sup ©} is an effective tool in reproducing the most significant phenomena of He-FUS3 system, providing relevant insight supporting future experimental campaigns. The post-test analysis phase will be, of course, fundamental for the qualification of a consistent numerical model.

  19. Development of thermal-hydraulic system analysis code SSC-K for pool-type liquid metal reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Doo; Kwon, Y.M.; Suk, S.D.; Chang, W.P.; Hahn, D.H

    1999-03-01

    The Supper System Code of KAERI (SSC-K) is a best-estimate system code for analyzing an variety of off-normal or accident of a pool type design. It is developed at KAERI on the basis of SSC-L developed at BNL to analyze pool-type LMR transients. Because of inherent difference between th pool and loop design, the major modefications of SSC-L is required for the safety analysis of KALIMER. The major difference between KALIMER and general loop type LMRs exists in the primary heat transport system. In KALIMER, all of the essential components consisted of the primary heat transport system are located within the reactor vessel. This is contrast to the loop type LMRs, in which all the primary components are connected via piping to form loops attached externally to the reactor vessel. KALIMER has only one cover gas space. This eliminates the need for separate cover gas systems over liquid level in pump tanks and upper plenum. Since the sodium in hot pool is separated from cold p