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Sample records for lstf cold-leg small-break

  1. An analytical investigation of cold leg small break accidents of the ATLAS facility

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    Kim, Yeon-Sik, E-mail: yskim3@kaeri.re.kr; Cho, Seok; Choi, Ki-Yong

    2015-09-15

    A previous parametric study of the direct vessel injection (DVI) line breaks was re-evaluated to see its applicability to that of the cold leg (CL) pipe breaks in advanced thermal-hydraulic test loop for accident simulation (ATLAS). Evaluation results of the tests and analyses for the major parameters, e.g., the pressurizer (PZR) pressure, downcomer water level, collapsed core water level, and clad temperature, were compared for four different CL pipe break scenarios. The overall trends of the major parameters showed reasonable behaviors between the tests and analyses. The clad temperature showed conservative behaviors in the analyses using the suggested options. The suggested counter-current flow limit (CCFL) options for the fuel alignment plate (FAP) and cross-over legs (COLs) can be applicable to any small-break loss-of-coolant accident (SBLOCA) scenario for the CL pipe and DVI line breaks in the ATLAS tests.

  2. Comparison of CATHARE results with the experimental results of cold leg intermediate break LOCA obtained during ROSA-2/LSTF test 7

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    Mazgaj Piotr

    2016-01-01

    Full Text Available Thermal-hydraulic analysis is a key part in support of regulatory work and nuclear power plant design and operation. In the field of Loss Of Coolant Accident, evolutions of the regulations are discussed in various countries taking into account the very unlikely character of a double-ended guillotine break and questioning the necessity to study such an event with Design Basis Conditions assumptions. As a consequence, the consideration of intermediate size piping rupture becomes more and more important. The paper presents the modeling of the Test Facility ROSA-2/LSTF in the calculation code CATHARE 2.V2.5. OECD/NEA ROSA-2 Project Test 7 was conducted with the Large Scale Test Facility on June 14, 2012. The experiment simulated the thermal-hydraulic responses during a PWR 13% cold leg Intermediate Break Loss Of Coolant Accident (IBLOCA. The break was simulated by a cold leg upwardly mounted long break nozzle. The facility and the experiment conditions are modeled in CATHARE. The vessel is modeled by using a 3D module. A thermal-hydraulic analysis is conducted and the obtained results are subsequently compared with the experimental results from ROSA-2/LSTF Test 7. Evaluation of the differences between experimental and calculated results is discussed.

  3. Cold-Leg Small Break LOCA Analysis of APR1400 Plant Using a SPACE/sEM Code

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    Lim, Sang Gyu; Lee, Suk Ho; Yu, Keuk Jong; Kim, Han Gon; Lee, Jae Yong [Central Research Institute, KHNP, Ltd., Daejeon (Korea, Republic of)

    2013-10-15

    The Small Break Loss-of-Coolant Accident (SBLOCA) evaluation methodology (EM) for APR1400, called sEM, is now being developed using SPACE code. SPACE/sEM is to set up a conservative evaluation methodology in accordance with appendix K of 10 CFR 50. Major required and acceptable features of the evaluation models are described as below. - Fission product decay : 1.2 times of ANS97 decay curve - Critical flow model : Henry-Fauske Moody two phase critical flow model - Metal-Water reaction model : Baker-Just equation - Critical Heat Flux (CHF) : B and W, Barnett and Modified Barnett correlation - Post-CHF : Groeneveld 5.7 film boiling correlation A series of test matrix is established to validate SPACE/sEM code in terms of major SBLOCA phenomena, e.g. core level swelling and boiling, core heat transfer, critical flow, loop seal clearance and their integrated effects. The separated effect tests (SETs) and integrated effect tests (IETs) are successfully performed and these results shows that SPACE/sEM code has a conservatism comparing with experimental data. Finally, plant calculations of SBLOCA for APR1400 are conducted as described below. - Break location sensitivity : DVI line, hot-leg, cold-leg, pump suction leg. - Break size spectrum : 0.4ft{sup 2}∼0.02ft{sup 2}(DVI) 0.5ft{sup 2}∼0.02ft{sup 2}(hot-leg, cold-leg, pump suction leg) This paper deals with break size spectrum analysis of cold-leg break accidents. Based on the calculation results, emergency core cooling system (ECCS) performances of APR1400 and typical SBLOCA phenomena can be evaluated. Cold-leg SBLOCA analysis for APR1400 is performed using SPACE/sEM code under harsh environment condition. SPACE/sEM code shows the typical SBLOCA behaviors and it is reasonably predicted. Although SPACE/sEM code has conservative models and correlations based on appendix K of 10 CFR 50, PCT does not exceed the requirement (1477 K). It is concluded that ECCS in APR1400 has a sufficient performance in cold-leg SBLOCA.

  4. ROSA/LSTF experiment report for RUN SB-CL-24 repeated core heatup phenomena during 0.5% cold leg break LOCA

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    Suzuki, Mitsuhiro; Anoda, Yoshinari [Department of Reactor Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

    2000-03-01

    A small break loss-of-coolant accident (SBLOCA) in a Westinghouse-type four-loop PWR was simulated in an experiment (SB-CL-24) conducted at the Large-Scale Test Facility (LSTF) with an intention to study repeated core heatup during a long-term cooldown process. The experiment was conducted on February 28, 1990 with specified test conditions including failure assumptions both on the high pressure injection (HPI) and the auxiliary feedwater systems, and the intentional secondary system depressurization as an operator action. The secondary depressurization contributed to promote the primary depressurization and the actuation of accumulator injection system (AIS). A temporary core heatup was observed in each of three loopseal clearing (LSC) processes. A significant core heatup occurred in the following boil-off process after loss of the secondary coolant mass and the AIS termination due to increase of the primary pressure. By additional opening of the pressurizer relief valves and safety valves, the primary pressure rapidly decreased to result in the low pressure injection (LPI) which cooled the heated core. This report summarizes results of the experiment (SB-CL-24) in addition to typical responses of some accident indication systems including the core exit thermocouples (CETs) and the water level meters in the primary system. (author)

  5. RELAP5 Analyses of ROSA/LSTF Experiments on AM Measures during PWR Vessel Bottom Small-Break LOCAs with Gas Inflow

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    Takeshi Takeda

    2014-01-01

    Full Text Available RELAP5 code posttest analyses were performed on ROSA/LSTF experiments that simulated PWR 0.2% vessel bottom small-break loss-of-coolant accidents with different accident management (AM measures under assumptions of noncondensable gas inflow and total failure of high-pressure injection system. Depressurization of and auxiliary feedwater (AFW injection into the secondary-side of both steam generators (SGs as the AM measures were taken 10 min after a safety injection signal. The primary depressurization rate of 55 K/h caused rather slow primary depressurization being obstructed by the gas accumulation in the SG U-tubes after the completion of accumulator coolant injection. Core temperature excursion thus took place by core boil-off before the actuation of low-pressure injection (LPI system. The fast primary depressurization by fully opening the relief valves in both SGs with continuous AFW injection led to long-term core cooling by the LPI actuation even under the gas accumulation in the SG U-tubes. The code indicated remaining problems in the predictions of break flow rate during two-phase flow discharge period and primary pressure after the gas inflow. Influences of the primary depressurization rate with continuous AFW injection onto the long-term core cooling were clarified by the sensitivity analyses.

  6. Assessment of RELAP/MOD3 using BETHSY 6.2TC 6-inch cold leg side break comparative test

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    Chung, Young-Jong; Jeong, Jae-Jun; Chang, Won-Pyo; Kim, Dong-Su [Korea Atomic Energy Research Institute, Yusung, Taejon (Korea, Republic of)] [and others

    1996-10-01

    This report presents the results of the RELAP5/MOD3 Version 7j assessment on BETHSY 6.2TC. BETHSY 6.2TC test corresponding to a six inch cold leg break LOCA of the Pressurizer Water Reactor(PWR). The primary objective of the test was to provide reference data of two facilities of different scales (BETHSY and LSTF facility). On the other hand, the present calculation aims at analysis of RELAP5/N4OD3 capability on the small break LOCA simulation, The results of calculation have shown that the RELAP5/MOD3 reasonably predicts occurrences as well as trends of the major phenomena such as primary pressure, timing of loop seal clearing, liquid hold up, etc. However, some disagreements also have been found in the predictions of loop seal clearing, collapsed core water level after loop seal clearing, and accumulator injection behaviors. For better understanding of discrepancies in same predictions, several sensitivity calculations have been performed as well. These include the changes of two-phase discharge coefficient at the break junction and some corrections of the interphase drag term. As result, change of a single parameter has not improved the overall predictions and it has been found that the interphase drag model has still large uncertainties.

  7. ROSA/LSTF Tests and RELAP5 Posttest Analyses for PWR Safety System Using Steam Generator Secondary-Side Depressurization against Effects of Release of Nitrogen Gas Dissolved in Accumulator Water

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    Takeshi Takeda

    2016-01-01

    Full Text Available Two tests related to a new safety system for a pressurized water reactor were performed with the ROSA/LSTF (rig of safety assessment/large scale test facility. The tests simulated cold leg small-break loss-of-coolant accidents with 2-inch diameter break using an early steam generator (SG secondary-side depressurization with or without release of nitrogen gas dissolved in accumulator (ACC water. The SG depressurization was initiated by fully opening the depressurization valves in both SGs immediately after a safety injection signal. The pressure difference between the primary and SG secondary sides after the actuation of ACC system was larger in the test with the dissolved gas release than that in the test without the dissolved gas release. No core uncovery and heatup took place because of the ACC coolant injection and two-phase natural circulation. Long-term core cooling was ensured by the actuation of low-pressure injection system. The RELAP5 code predicted most of the overall trends of the major thermal-hydraulic responses after adjusting a break discharge coefficient for two-phase discharge flow under the assumption of releasing all the dissolved gas at the vessel upper plenum.

  8. IIST small break LOCA experiments with passive core cooling injection

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    Chang, C.-J. [Nuclear Engineering Division, Institute of Nuclear Energy Research, P.O. Box 3-3, Longtan, Taiwan 325 (China)]. E-mail cjchang@iner.gov.tw; Lee, C.-H. [Nuclear Engineering Division, Institute of Nuclear Energy Research, P.O. Box 3-3, Longtan, Taiwan 325 (China); Hong, W.-T. [Nuclear Engineering Division, Institute of Nuclear Energy Research, P.O. Box 3-3, Longtan, Taiwan 325 (China); Wang, Lance L.C. [Lungmen Project Quality Supervisory and Directory Committee, Taiwan Power Company, 62, Yeh-Hai St., Kung-Liao, Taipei County, Taiwan 238 (China)

    2006-01-15

    The purpose of this study is to evaluate the performance of a passive core cooling system (PCCS) with passive injection during the cold-leg small break loss-of-coolant accidents (SBLOCAs) experiments conducted at the Institute of Nuclear Energy Research (INER) Integral System Test (IIST) facility. Four tests were performed simulating break sizes of 0.2-2% (approximately corresponding to 1.25-4'' breaks for a referenced nuclear power plant) at cold-leg for assessing the PCCS capability in accident management. The key thermal-hydraulic phenomena to core heat removal for PCCS are observed and discussed. The experimental results show that the PCCS has successfully provided a continuous removal of core heat and a long term core cooling can be reached for all cases of SBLOCA.

  9. Post-test analysis of the ROSA/LSTF and PKL counterpart test

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    Carlos, S., E-mail: scarlos@iqn.upv.es [Departament d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera, 14, València Spain (Spain); Querol, A., E-mail: anquevi@iqn.upv.es [Departament d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera, 14, València Spain (Spain); Instituto de Seguridad Industrial, Radiofísica y Medioambiental, Universitat Politècnica de València, Camí de Vera, 14, València (Spain); Gallardo, S., E-mail: sergalbe@iqn.upv.es [Departament d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera, 14, València Spain (Spain); Instituto de Seguridad Industrial, Radiofísica y Medioambiental, Universitat Politècnica de València, Camí de Vera, 14, València (Spain); Sanchez-Saez, F., E-mail: frasansa@etsii.upv.es [Departament d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera, 14, València Spain (Spain); and others

    2016-02-15

    Highlights: • TRACE modelization for PKL and ROSA/LSTF installations. • Secondary-side depressurization as accident management action. • CET vs PCT relation. • Analysis of differences in the vessel models. - Abstract: Experimental facilities are scaled models of commercial nuclear power plants, and are of great importance to improve nuclear power plants safety. Thus, the results obtained in the experiments undertaken in such facilities are essential to develop and improve the models implemented in the thermal-hydraulic codes, which are used in safety analysis. The experiments and inter-comparisons of the simulated results are usually performed in the frame of international programmes in which different groups of several countries simulate the behaviour of the plant under the accidental conditions established, using different codes and models. The results obtained are compared and studied to improve the knowledge on codes performance and nuclear safety. Thus, the Nuclear Energy Agency (NEA), in the nuclear safety work area, auspices several programmes which involve experiments in different experimental facilities. Among the experiments proposed in NEA programmes, one on them consisted of performing a counterpart test between ROSA/LSTF and PKL facilities, with the main objective of determining the effectiveness of late accident management actions in a small break loss of coolant accident (SBLOCA). This study was proposed as a result of the conclusion obtained by the NEA Working Group on the Analysis and Management of Accidents, which analyzed different installations and observed differences in the measurements of core exit temperature (CET) and maximum peak cladding temperature (PCT). In particular, the transient consists of a small break loss of coolant accident (SBLOCA) in a hot leg with additional failure of safety systems but with accident management measures (AM), consisting of a fast secondary-side depressurization, activated by the CET. The paper

  10. ROSA-V large scale test facility (LSTF) system description for the third and fourth simulated fuel assemblies

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    Suzuki, Mitsuhiro; Nakamura, Hideo; Ohtsu, Iwao [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    2003-03-01

    The Large Scale Test Facility (LSTF) is a full-height and 1/48 volumetrically scaled test facility of the Japan Atomic Energy Research Institute (JAERI) for system integral experiments simulating the thermal-hydraulic responses at full-pressure conditions of a 1100 MWe-class pressurized water reactor (PWR) during small break loss-of-coolant accidents (SBLOCAs) and other transients. The LSTF can also simulate well a next-generation type PWR such as the AP600 reactor. In the fifth phase of the Rig-of-Safety Assessment (ROSA-V) Program, eighty nine experiments have been conducted at the LSTF with the third simulated fuel assembly until June 2001, and five experiments have been conducted with the newly-installed fourth simulated fuel assembly until December 2002. In the ROSA-V program, various system integral experiments have been conducted to certify effectiveness of both accident management (AM) measures in beyond design basis accidents (BDBAs) and improved safety systems in the next-generation reactors. In addition, various separate-effect tests have been conducted to verify and develop computer codes and analytical models to predict non-homogeneous and multi-dimensional phenomena such as heat transfer across the steam generator U-tubes under the presence of non-condensable gases in both current and next-generation reactors. This report presents detailed information of the LSTF system with the third and fourth simulated fuel assemblies for the aid of experiment planning and analyses of experiment results. (author)

  11. Analysis of the ATLAS Cold Leg Top-Slot Break Experiment Using the MARS Code

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    Ha, T. W.; Jeong, J. J. [Pusan National University, Busan (Korea, Republic of)

    2016-10-15

    During a small-break loss of coolant accident (SBLOCA) or intermediate-break loss of coolant accident (IBLOCA) in a PWR, such as the APR1400, the steam volume in the reactor vessel upper plenum may continue to expand until the liquid phase in the horizontal intermediate legs is released, called loop seal clearing (LSC), due to the increase of the pressure in the upper plenum. A domestic standard problem (DSP) exercise using the ATLAS facility was promoted in order to transfer the database to domestic nuclear industries. For 4th DSP (DSP-04), the ATLAS cold leg top-slot break experiment was postulated. For the DSP-04, main concerns are to predict the LSC and LSR having a significantly effect on the behavior of the system under long term cooling. In this study, we simulated the ATLAS cold leg top-slot break experiment using the MARS code and the predicted LSC and LSR are compared to experimental results. The LTC-CL-04R was simulated using the MARS code. Most of the predicted results agree well with the experimental data. However, the timing of LSC and LSR is slightly different from each other and, thus, the behavior of the primary system is slightly different. The core heat up was not observed in the experiment and the calculation.

  12. Small Break Air Ingress Experiment

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    Chang Oh; Eung Soo Kim

    2011-09-01

    The small break air-ingress experiment, described in this report, is designed to investigate air-ingress phenomena postulated to occur in pipes in a very high temperature gas-cooled reactor (VHTRs). During this experiment, air-ingress rates were measured for various flow and break conditions through small holes drilled into a pipe of the experimental apparatus. The holes were drilled at right angles to the pipe wall such that a direction vector drawn from the pipe centerline to the center of each hole was at right angles with respect to the pipe centerline. Thus the orientation of each hole was obtained by measuring the included angle between the direction vector of each hole with respect to a reference line anchored on the pipe centerline and pointing in the direction of the gravitational force. Using this reference system, the influence of several important parameters on the air ingress flow rate were measured including break orientation, break size, and flow velocity . The approach used to study the influence of these parameters on air ingress is based on measuring the changes in oxygen concentrations at various locations in the helium flow circulation system as a function of time using oxygen sensors (or detectors) to estimate the air-ingress rates through the holes. The test-section is constructed of a stainless steel pipe which had small holes drilled at the desired locations.

  13. Comparison of CATHARE results with the experimental results of cold leg intermediate break LOCA obtained during ROSA-2/LSTF test 7

    OpenAIRE

    2016-01-01

    Thermal-hydraulic analysis is a key part in support of regulatory work and nuclear power plant design and operation. In the field of Loss Of Coolant Accident, evolutions of the regulations are discussed in various countries taking into account the very unlikely character of a double-ended guillotine break and questioning the necessity to study such an event with Design Basis Conditions assumptions. As a consequence, the consideration of intermediate size piping rupture becomes more and more i...

  14. ANALISIS KONDISI TERAS REAKTOR DAYA MAJU AP1000 PADA KECELAKAAN SMALL BREAK LOCA

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    Andi Sofrany Ekariansyah

    2015-06-01

    , mixture level, temperatur kelongsong, small break LOCA, RELAP5.   ABSTRACT ANALYSIS ON THE CORE CONDITION OF AP1000 ADVANCED POWER REACTOR DURING SMALL BREAK LOCA. Accident due to the loss of coolant from the reactor boundary is an anticipated design basis event in the design of power reactor adopting the Generation II up to IV technology. Small break LOCA leads to more significant impact on safety compared to the large break LOCA as shown in the Three-Mile Island (TMI. The focus of this paper is the small break LOCA analysis on the Generation III+ advanced power reactor of AP1000 by simulating three different initiating events, which are inadvertent opening of Automatic Depressurization System (ADS, double-ended break on one of Direct Vessel Injection (DVI pipe, and 10 inch diameter split break on one of cold leg pipe. Methodology used is by simulating the events on the AP1000 model developed using RELAP5/SCDAP/Mod3.4. The impact analyzed is the core condition during the small break LOCA consisting of the mixture level occurrence and the fuel cladding temperature transient. The results show that the mixture level for all small break LOCA events are above the active core height, which indicates no core uncovery event. The development of the mixture level affect the fuel cladding temperature transient, which shows a decreasingly trend after the break, and the effectifeness of core cooling. Those results are identical with the results of other code of NOTRUMP. The resulted core cooling is also due to the function of coolant injection from passive safety feature, which is unique in the AP1000 design. In overall, the result of analysis shows that the AP1000 model developed by the RELAP5 can be used for analysis of design basis accident considered in the AP1000 advanced power reactor. Keywords: analysis, mixture level, fuel cladding temperature, small break LOCA, RELAP5.

  15. International Standard Problems and Small Break Loss-of-Coolant Accident (SBLOCA

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    N. Aksan

    2008-01-01

    Full Text Available Best-estimate thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants and also used in the design of advance reactors. Evaluation of the capabilities and the performance of these codes can be accomplished by comparing the code predictions with measured experimental data obtained on different test facilities. OECD/NEA Committee on the Safety of Nuclear Installations (CSNI has promoted, over the last twenty-nine years, some forty-eight international standard problems (ISPs. These ISPs were performed in different fields as in-vessel thermal-hydraulic behaviour, fuel behaviour under accident conditions, fission product release and transport, core/concrete interactions, hydrogen distribution and mixing, containment thermal-hydraulic behaviour. 80% of these ISPs were related to the working domain of principal working group no.2 on coolant system behaviour (PWG2 and were one of the major PWG2 activities for many years. A global review and synthesis on the contribution that ISPs have made to address nuclear reactor safety issues was initiated by CSNI-PWG2 and an overview on the subject of small break LOCA ISPs is given in this paper based on a report prepared by a writing group. In addition, the relevance of small break LOCA in a PWR with relation to nuclear reactor safety and the reorientation of the reactor safety program after TMI-2 accident are shortly summarized. The experiments in four integral test facilities, LOBI, SPES, BETHSY, ROSA IV/LSTF and the recorded data during a steam generator tube rupture transient in the DOEL-2 PWR (Belgium were the basis of the five small break LOCA related ISP exercises, which deal with the phenomenon typical of small break LOCAs in Western design PWRs. Some lessons learned from these small break LOCA ISPs are identified in relation to code deficiencies and capabilities, progress in the code capabilities, possibility of scaling, and various additional aspects

  16. Scaling of a break small in the hot field of the LSTF installation to a nuclear power plant; Escalado de una rotura pequena en la rama caliente de la instacions LSTF a una central nuclear

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    Querol, A.; Gallardo, S.; Verdu, G.

    2014-07-01

    The objective of this work is to study if the phenomenology observed during in one break small in the hot branch of a small scale, such as Large Scale Test installation Facility (LSTF), can be extrapolated to a real plant. To do so, is considered the experiment 1-2 of the OECD/NEA rose, which simulates a break in the hot field of 1%, and the code project Thermo-hydraulic TRACE5 to reproduce the LSTF facility and a real plant and thus, power compare the results. They have been considered two models for the actual plant: a scaled model of the LSTF facility while preserving the relationship of power and volume of original LSTF and a typical model Westinghouse PWR of 3 loops. The results obtained show that the scaling LSTF and LSTF installation models reproduce the same behavior during the transient. However, with the Westinghouse model 3 ties there are important differences. The most important thing is that natural circulation does not play properly. The effect of the nodalization of U-tubes and the vessel has been in improving outcomes. The results obtained show that the nodalization of U-tubes affects the reproduction of natural circulation, while the nodalization of the vessel does not have a relevant effect. (Author)

  17. Sensitivity Analyses in Small Break LOCA with HPI-Failure: Effect of Break-Size in Secondary-Side Depressurization

    Science.gov (United States)

    Kinoshita, Ikuo; Torige, Toshihide; Yamada, Minoru

    2014-06-01

    In the case of total failure of the high pressure injection (HPI) system following small break loss of coolant accident (SBLOCA) in pressurized water reactor (PWR), the break size is so small that the primary system does not depressurize to the accumulator (ACC) injection pressure before the core is uncovered extensively. Therefore, steam generator (SG) secondary-side depressurization is necessary as an accident management in order to grant accumulator system actuation and core reflood. A thermal-hydraulic analysis using RELAP5/MOD3 was made on SBLOCA with HPI-failure for Oi Units 3/4 operated by Kansai Electoric Power Co., which are conventional 4 loop PWR plants. The effectiveness of SG secondary-side depressurization procedure was investigated for the real plant design and operational characteristics. The sensitivity analyses using RELAP5/MOD3.2 showed that the accident management was effective for a wide range of break sizes, various orientations and positions. The critical break can be 3 inch cold-leg bottom break.

  18. Simulation and Analysis of Small Break LOCA for AP1000 Using RELAP5-MV and Its Comparison with NOTRUMP Code

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    Eltayeb Yousif

    2017-01-01

    Full Text Available Many reactor safety simulation codes for nuclear power plants (NPPs have been developed. However, it is very important to evaluate these codes by testing different accident scenarios in actual plant conditions. In reactor analysis, small break loss of coolant accident (SBLOCA is an important safety issue. RELAP5-MV Visualized Modularization software is recognized as one of the best estimate transient simulation programs of light water reactors (LWR. RELAP5-MV has new options for improved modeling methods and interactive graphics display. Though the same models incorporated in RELAP5/MOD 4.0 are in RELAP5-MV, the significant difference of the latter is the interface for preparing the input deck. In this paper, RELAP5-MV is applied for the transient analysis of the primary system variation of thermal hydraulics parameters in primary loop under SBLOCA in AP1000 NPP. The upper limit of SBLOCA (10 inches is simulated in the cold leg of the reactor and the calculations performed up to a transient time of 450,000.0 s. The results obtained from RELAP5-MV are in good agreement with those of NOTRUMP code obtained by Westinghouse when compared under the same conditions. It can be easily inferred that RELAP5-MV, in a similar manner to RELAP5/MOD4.0, is suitable for simulating a SBLOCA scenario.

  19. Flow regimes and heat transfer modes identification in ANGRA 2 core, during small break in the primary loop with area of 100 cm{sup 2}, simulated with RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Eduardo M.; Sabundjian, Gaiane, E-mail: gdgian@ipen.br, E-mail: borges.em@hotmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    Identifying the flow regimes and the heat transfer modes is important for the analysis of accidents such as the Loss-of-Coolant Accident (LOCA). The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used in the RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 100cm{sup 2}-rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR - A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. (author)

  20. Comparative Experiments to Assess the Effects of Accumulator Nitrogen Injection on Passive Core Cooling During Small Break LOCA

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    Li Yuquan

    2017-02-01

    Full Text Available The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA, the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility—the advanced core-cooling mechanism experiment (ACME—was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA transient. Two comparison test groups—a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI line break—were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the

  1. Core thermal hydraulic behavior during the reflood phase of cold-leg LBLOCA experiments using the ATLAS test facility

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Seok; Park, Hyun Sik; Choi, Ki Yong; Kang, Kyoung Ho; Baek, Won Pil; Kim, Yeon Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-12-15

    Several experimental tests to simulate a reflood phase of a cold-leg LBLOCA of the APR1400 have been performed using the ATLAS facility. This paper describes the related experimental results with respect to the thermal-hydraulic behavior in the core and the system-core interactions during the reflood phase of the cold-leg LBLOCA conditions. The present descriptions will be focused on the LB-CL-09, LB-CL-11, LB-CL-14, and LB-CL-15 tests performed using the ATLAS. The LB-CL-09 is an integral effect test with conservative boundary condition; the LB-CL-11 and -14 are integral effect tests with realistic boundary conditions, and the LB-CL-15 is a separated effect test. The objectives of these tests are to investigate the thermal-hydraulic behavior during an entire reflood phase and to provide reliable experimental data for validating the LBLOCA analysis methodology for the APR1400. The initial and boundary conditions were obtained by applying scaling ratios to the MARS simulation results for the LBLOCA scenario of the APR1400. The ECC water flow rate from the safety injection tanks and the decay heat were simulated from the start of the reflood phase. The simulated core power was controlled to be 1.2 times that of the ANS-73 decay heat curve for LB-CL-09 and 1.02 times that of the ANS-79 decay curve for LB-CL-11, -14, and -15. The simulated ECC water flow rate from the high pressure safety injection pump was 0.32 kg/s. The present experimental data showed that the cladding temperature behavior is closely related to the collapsed water level in the core and the downcomer

  2. Robinson 2 reactor vessel: pressurized thermal shock analysis for a small-break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Marston, T.; Griesbach, T.; Chao, J.; Chexal, B.; Norris, D.; Nickell, B.; Layman, B.

    1984-08-01

    A best-estimate Pressurized Thermal Shock (PTS) analysis was performed for a three-inch diameter hot-leg small-break loss-of-coolant accident for the Robinson 2 plant. This plant specific analysis was performed using EPRI's linked set of codes for PTS analysis. The analysis shows that with the H.B. Robinson 2 reactor pressure vessel, a hot-leg small-break loss-of-coolant accident does not pose a significant health or safety concern to the public for at least 40 years of operation.

  3. CATHARE2 calculation of SPE-3 test small break loca on PMK facility

    Energy Technology Data Exchange (ETDEWEB)

    Laugier, E.; Radet, J. [Institut de Protection et de Surete Nucleaire, Cadarache (France)

    1995-09-01

    Bind and post test calculations with CATHARE2 have been performed concerning the SPE-4 exercise organized under the auspices of IAEA on the hungarian PMK-2 facility, a one loop scaled model of VVER 440/213 Nuclear Power Plant. The SPE-4 test is a cold leg SBLOCA associated to a {open_quotes}bleed and feed{close_quotes} procedure applied in the secondary circuit. The present paper is devoted to the analysis of the post test calculation. For the first part of the transient (until the end of the SIT activations), the primary and secondary pressures are rather well predicted, leading to a good agreement with the experimental trips, as scram, flow coast down, SIT beginning and end of activation. Nevertheless, some discrepancy with the experiment may be due to an over prediction of the thermal exchanges from the primary to the secondary circuits. For the second part of the transient, the predicted primary circuit repressurization is shifted after the SITs are off, while in the experiment this event immediately follows the end of SIT activation. The delay in the calculation leads to underpredict primary and secondary pressures, thus anticipating the timing of events, such as LPIS and emergency feedwater activation.

  4. Analysis of a small break loss-of-coolant accident of pressurized water reactor by APROS

    Energy Technology Data Exchange (ETDEWEB)

    Al-Falahi, A. [Helsinki Univ. of Technology, Espoo (Finland); Haennine, M. [VTT Energy, Espoo (Finland); Porkholm, K. [IVO International, Ltd., Vantaa (Finland)

    1995-09-01

    The purpose of this paper is to study the capability of APROS (Advanced PROcess Simulator) code to simulate the real plant thermal-hydraulic transient of a Small Break Loss-Of-Coolant Accident (SBLOCA) of Loss-Of-Fluid Test (LOFT) facility. The LOFT is a scaled model of a Pressurized Water Reactor (PWR). This work is a part of a larger validation of the APROS thermal-hydraulic models. The results of SBLOCA transient calculated by APROS showed a reasonable agreement with the measured data.

  5. Small-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Boyack, B.E.; Steiner, J.L.; Harmony, S.C. [Los Alamos National Lab., Albuquerque, NM (United States)] [and others

    1995-09-01

    The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions having a very low probability of occurrence.

  6. TASS/SMR code improvement for small break LOCA applicability at an integral type reactor, SMART

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Young-Jong, E-mail: chung@kaeri.re.kr; Kim, Soo-Hyung; Lim, Sung-Won; Bae, Kyoo-Hwan

    2015-12-15

    Highlights: • SMART adopts a passive system to enhance its safety. • TASS/SMR code is developed to analyze thermal hydraulic phenomena of the SMART plant. • Improved TASS/SMR code predicts well the results of the OSU-MASLWR total-loss-of-feedwater test. - Abstract: Small reactors are a suitable option for nuclear system deployment in developing countries or non-electrical applications for various facilities. SMART is one of the small integral type reactors to apply flexibly local power demands or sea water desalination. A thermal hydraulic analysis code, TASS/SMR, having SMART specific models, was developed to simulate thermal hydraulic phenomena of the SMART plant. The improved TASS/SMR code predicts well the system behaviors under two-phase conditions compared with the OSU-MASLWR experimental results. A small break LOCA simulation of the SMART plant is improved a void distribution, a break flow, and a collapsed water level in the core.

  7. Hydrogen and steam distribution following a small-break LOCA in large dry containment

    Institute of Scientific and Technical Information of China (English)

    DENG Jian; CAO Xuewu

    2007-01-01

    The hydrogen deflagration is one of the major risk contributors to threaten the integrity of the containment in a nuclear power plant, and hydrogen control in the case of severe accidents is required by nuclear regulations.Based on the large dry containment model developed with the integral severe-accident analysis tool, a small-break loss-of-coolant-accident (LOCA) without HPI, LPI, AFW and containment sprays, leading to the core degradation and large hydrogen generation, is calculated. Hydrogen and steam distribution in containment compartments is investigated. The analysis results show that significant hydrogen deflagration risk exits in the reactor coolant pump (RCP)compartment and the cavity during the early period, if no actions are taken to mitigate the effects of hydrogen accumulation.

  8. Experimental simulation of asymmetric heat up of coolant channel under small break LOCA condition for PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Yadav, Ashwini K., E-mail: ashwinikumaryadav@gmail.com [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667 (India); Majumdar, P., E-mail: pmajum@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Kumar, Ravi, E-mail: ravikfme@iitr.ernet.in [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667 (India); Chatterjee, B., E-mail: barun@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Gupta, Akhilesh, E-mail: akhilfme@iitr.ernet.in [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667 (India); Mukhopadhyay, D., E-mail: dmukho@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2013-02-15

    Highlights: ► Circumferential temperature gradient of PT for asymmetric heat-up was 440 °C. ► At 2 MPa ballooning initiated at 450 °C and with strain rate of 0.0277%/s. ► At 4 MPa ballooning initiated at 390 °C and with strain rate of 0.0305%/s. ► At 4 MPa, PT ruptured under uneven strain and steep temperature gradient. ► Integrity of PT depends on internal pressure and magnitude of decay power. -- Abstract: During postulated small break loss of coolant accident (SBLOCA) for Pressurised Heavy Water Reactors (PHWRs) as well as for postulated SBLOCA coincident with loss of ECCS, a stratified flow condition can arise in the coolant channels as the gravitational force dominates over the low inertial flow arising from small break flow. A Station Blackout condition without operator intervention can also lead to stratified flow condition during a slow channel boil-off condition. For all these conditions the pressure remains high and under stratified flow condition, the horizontal fuel bundles experience different heat transfer environments with respect to the stratified flow level. This causes the bundle upper portion to get heated up higher as compared to the submerged portion. This kind of asymmetrical heating of the bundle is having a direct bearing on the circumferential temperature gradient of pressure tube (PT) component of the coolant channel. The integrity of the PT is important under normal conditions as well as at different accident loading conditions as this component houses the fuel bundles and serves as a coolant pressure boundary of the reactors. An assessment of PT is required with respect to different accident loading conditions. The present investigation aims to study thermo-mechanical behaviour of PT (Zr, 2.5 wt% Nb) under a stratified flow condition under different internal pressures. The component is subjected to an asymmetrical heat-up conditions as expected during the said situation under different pressure conditions which varies from 2

  9. Analysis of different containment models for IRIS small break LOCA, using GOTHIC and RELAP5 codes

    Energy Technology Data Exchange (ETDEWEB)

    Papini, Davide, E-mail: davide.papini@mail.polimi.i [Department of Energy, CeSNEF - Nuclear Engineering Division, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Grgic, Davor [Department of Power Systems, FER, University of Zagreb, Unska 3, 10000 Zagreb (Croatia); Cammi, Antonio; Ricotti, Marco E. [Department of Energy, CeSNEF - Nuclear Engineering Division, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy)

    2011-04-15

    Advanced nuclear water reactors rely on containment behaviour in realization of some of their passive safety functions. Steam condensation on containment walls, where non-condensable gas effects are significant, is an important feature of the new passive containment concepts, like the AP600/1000 ones. In this work the international reactor innovative and secure (IRIS) was taken as reference, and the relevant condensation phenomena involved within its containment were investigated with different computational tools. In particular, IRIS containment response to a small break LOCA (SBLOCA) was calculated with GOTHIC and RELAP5 codes. A simplified model of IRIS containment drywell was implemented with RELAP5 according to a sliced approach, based on the two-pipe-with-junction concept, while it was addressed with GOTHIC using several modelling options, regarding both heat transfer correlations and volume and thermal structure nodalization. The influence on containment behaviour prediction was investigated in terms of drywell temperature and pressure response, heat transfer coefficient (HTC) and steam volume fraction distribution, and internal recirculating mass flow rate. The objective of the paper is to preliminarily compare the capability of the two codes in modelling of the same postulated accident, thus to check the results obtained with RELAP5, when applied in a situation not covered by its validation matrix (comprising SBLOCA and to some extent LBLOCA transients, but not explicitly the modelling of large dry containment volumes). The option to include or not droplets in fluid mass flow discharged to the containment was the most influencing parameter for GOTHIC simulations. Despite some drawbacks, due, e.g. to a marked overestimation of internal natural recirculation, RELAP5 confirmed its capability to satisfactorily model the basic processes in IRIS containment following SBLOCA.

  10. Pre-test analysis of an integral effect test facility for thermal hydraulic similarities of 6 inches cold leg break and DVI line break using MARS-1D

    Energy Technology Data Exchange (ETDEWEB)

    Ah, D. J.; Park, H. S.; Choi, K. Y.; Kwon, T. S.; Baek, W. P. [KAERI, Taejon (Korea, Republic of)

    2002-05-01

    Pre-test analyses of a small-break loss-of-coolant accident (SBLOCA) and a DVI line break accident, have been performed for the integral effect test loop of Korea Atomic Energy Research Institute (KAERI-ITL), the construction of which will be started soon. The KAERI-ITL is being designed with a full-height and 1/310 volume scale based on the design features of the APR1400 (Korean Next Generation Reactor). Based on the same control logics and accident scenarios, the similarity between the KAERI-ITL and the prototype plant, APR1400, is evaluated using the MARS code. It is found that the KAERI-ITL and APR1400 have similar thermal hydraulic responses during the transient under the identical accident scenarios. It is also verified that the volume scaling law, applied to the design of the KAERI-ITL, gives reasonable results to keep the similarity between APR1400 and KAERI-ITL.

  11. Effect of fuel pin ballooning on the sub-channel thermal hydraulics during small break loca for Indian PHWRS

    Energy Technology Data Exchange (ETDEWEB)

    Mukhopadhyay, D.; Behera, G.H.; Bandopadhyay, S.K.; Gupta, S.K. [Bhabha Atomic Research Centre, Div. Reactor Safety, Bombay (India)

    2001-07-01

    Effect of fuel pin ballooning on the subchannel thermal-hydraulics during a small break (0.25%) located at the Reactor Inlet Feeder (RIF) has been studied for Indian PHWRs. The break leads to a low flow situation in the affected reactor channel along with delayed reactor trip. Higher power to flow ratio in the inner subchannels in comparison to outer subchannel of a 19 pin fuel bundle causes early 2-phase condition causing the flow to by pass from the inner ones to outer ones. This causes the fuel pins to experience different temperatures. Fuel pin ballooning causes reduction in the subchannel areas and further flow redistribution takes place. The transient subchannel thermal-hydraulic conditions along the reactor channel are very much different due to the power distribution and pressure drop. (authors)

  12. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B. [and others

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.

  13. Evaluation of a postulated loss of coolant accident (LOCA) due to a 160 cm{sup 2} break in a cold leg of Angra 2 nuclear power plant[LOCA; RELAP-5/MOD3.2.2g; nodalization]; Calculo do acidente postulado de perda de refrigerante por uma ruptura de 160 cm{sup 2} na perna fria da central nuclear Angra 2

    Energy Technology Data Exchange (ETDEWEB)

    Azevedo, Carlos Vicente Goulart de; Palmieri, Elcio Tadeu; Aronne, Ivan Dionysio [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)

    2002-07-01

    The development of a qualified full nodalization of Angra2 NPP for RELAP5/Mod 3.2.2 gamma, aiming at the evaluation of a comprehensive number of accidents and transients, thus providing suitable safety analysis support for licensing purposes, is being carried out within the framework of CNEN internal technical cooperation, involving some of its institutes (CDTN, IPEN and IEN) and the Reactors Coordination (CODRE). This work presents a simulation of a postulated Angra2 small cold leg break loss of coolant accident (SBLOCA). A 160 cm{sup 2} break is supposed to occur at one cold leg between the main coolant pump and the reactor vessel and is described in the Angra2 Final Safety Analysis Report, section 15.6.4.1.3.4. The simulation of several types of transients and accidents is necessary to verify the adequate performance of the modeled logic and systems. In general, the analysis of such and accident allows to demonstrate the safety Injection System performance and the reliable transition between the high pressure safety injection, the accumulator injection and the residual heat removal phases. Furthermore, it is assumed that some components are out of service due to fail or repair in order to make a conservative analysis. The results showed a compatible behavior of the molded systems and that the simulated Emergency Core Cooling System was able to provide sufficient cooling to avoid any damage to the core. (author)

  14. Post test calculation of the experiment `small break loss-of- coolant test` SBL-22 at the Finnish integral test facility PACTEL with the thermohydraulic code ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Lischke, W.; Vandreier, B. [Univ. for Applied Sciences, Zittau/Goerlitz (Germany). Dept. of Nuclear Technology

    1997-12-31

    At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory. 5 refs.

  15. Experimental Investigation on Small Break Loss of Coolant Accident for Direct Vessel Injection Line%DVI管小破口失水事故实验研究

    Institute of Scientific and Technical Information of China (English)

    彭传新; 张妍; 黄志刚; 昝元锋; 卓文彬; 闫晓

    2016-01-01

    在模块化小型反应堆非能动安全系统综合模拟实验装置上进行了压力容器直接注入(DVI)管小破口失水事故实验,研究了DVI管小破口失水事故过程中的热工水力现象和非能动安全系统运行特性。研究结果表明:模块化小型反应堆DVI管小破口失水事故中,非能动安全系统可对堆芯进行注水,有效导出堆芯衰变热量,保护堆芯安全。%T he small break loss of coolant accident (SBLOCA ) experiment for direct vessel injection (DVI ) line , w hich investigated the thermal‐hydraulic phenomena and the performances of passive safety system during the accident ,was performed on the passive safety system test facility for small modular reactor .The experimental results show that the passive safety system of small modular reactor can provide effective cool‐ant injection ,successful removal of core residual heat under the DVI line SBLOCA and protection to reactor core safety .

  16. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Directory of Open Access Journals (Sweden)

    Hwang Bae

    2017-08-01

    Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  17. Condensation during gravity driven ECC: Experiments with PACTEL

    Energy Technology Data Exchange (ETDEWEB)

    Munther, R.; Kalli, H. [Lappeenranta Univ. of Technology (Finland); Kouhia, J. [Technical Research Centre of Finland, Lappeenranta (Finland)

    1995-09-01

    This paper provides the results of the second series of gravity driven emergency core cooling (ECC) experiments with PACTEL (Parallel Channel Test Loop). The simulated accident was a small break loss-of-coolant accident (SBLOCA) with a break in a cold leg. The ECC flow was provided from a core makeup tank (CMT) located at a higher elevation than the main part of the primary system. The CMT was pressurized with pipings from the pressurizer and a cold leg. The tests indicated that steam condensation in the CMT can prevent ECC and lead to core uncovery.

  18. Analysis of system thermal hydraulic responses for passive safety injection experiment at ROSA-IV Large Scale Test Facility. Using JAERI modified version of RELAP5/MOD2 code

    Energy Technology Data Exchange (ETDEWEB)

    Asaka, Hideaki; Yonomoto, Taisuke; Kukita, Yutaka (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment)

    1994-12-01

    An experiment was conducted at the ROSA-IV/Large Scale Test Facility (LSTF) on the performance of a gravity-driven emergency core coolant (ECC) injection system attached to a pressurized water reactor (PWR). Such a gravity-driven injection system, though not used in the current-generation PWRs, is proposed for future reactor designs. The experiment was performed to identify key phenomena peculiar to the operation of a gravity injection system and to provide data base for code assessment against such phenomena. The simulated injection system consisted of a tank which was initially filled with cold water of the same pressure as the primary system. The tank was connected at its top and bottom, respectively, to the cold leg and the vessel downcomer. The injection into the downcomer was driven primarily by the static head difference between the cold water in the tank and the hot water in the pressure balance line (PBL) connecting the cold leg to the tank top. The injection flow was oscillatory after the flow through the PBL became two-phase flow. The experiment was post-test analyzed using a JAERI modified version of the RELAP5/MOD2 code. The code calculation simulated reasonably well the system responses observed in the experiment, and suggested that the oscillations in the injection flow was caused by oscillatory liquid holdup in the PBL connecting the cold leg to tank top. (author).

  19. Thermal and fluid mixing in a 1/2-scale test facility. Volume 2. Data report. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Dolan, F.X.; Valenzuela, J.A.

    1985-07-01

    This report presents data from an experimental study of fluid mixing in a 1/2-scale model of the cold leg, downcomer, lower plenum, pump simulator, and loop seal typical of a Westinghouse Pressurized Water Reactor. The tests were transient cooldown tests in that they simulated an extreme condition of Small Break Loss of Coolant Accident (SBLOCA) during which cold High Pressure Injection (HPI) fluid is injected into stagnant, hot primary fluid with complete loss of natural circulation in the loop. Extensive temperature, velocity, and heat transfer coefficient data are presented at two cold leg Froude numbers: 0.052 and 0.076. The 1/2-scale data are compared with earlier data from a 1/5-scale, geometrically similar facility to assess scaling principles.

  20. Scientific design of the test facility for the KNGR DVI line small break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Byong Jo; Park, Choon Kyung; Jun, Hyung Gil; Cho, Seok; Kwon, Tae Soon; Song, Chul Hwa; Kim, Jung Taek

    1999-03-01

    Scientific design of the experimental facility (OASIS) for the KNGR (Korea Next Generation Reactor) DVI line SB-LOCA simulation is carried out. Main purpose of the OASIS is to produce thermal-hydraulic data base for determining the best location of the DVI (Direct Vessel Injection) injection nozzle of the KNGR as well as verifying its design performance in view of the ECCS (Emergency Core Cooling System) effectiveness. The experimental facility is designed based on the Ishii's three-level scaling law. The facility has 1/4 height and 1/341 area scaling ratio. It corresponds to the volume scale of 1/1364. The power scaling is 1/682 and the system pressure is prototypic. The OASIS consists of a core, a downcomer, two steam generators, two pump simulators, a break simulator, a collection tank, primary piping as well as a circulation pump for initial test condition. Each component is designed based on the Ishill's global scaling and boundary flow scaling of mass, energy and momentum. In addition, local phenomena scaling is carried out for the design of major components to preserve key local phenomena in each component. Most of the key phenomena are well preserved in the OASIS. However, the local scaling analysis shows that distortions of the void fraction and mixture level can not be avoided in the core. It comes from the basic features of the Ishill's scaling law in case of the reduced-height simulation. However, it is expected that these distortions will be analyzed properly by a best estimate system analysis code. (Author). 22 refs., 20 tabs., 25 figs.

  1. Experimental and numerical investigation of coolant mixing in a model of reactor pressure vessel down-comer and in cold leg inlets

    Directory of Open Access Journals (Sweden)

    Hutli Ezddin

    2017-01-01

    Full Text Available Thermal fatigue and pressurized thermal shock phenomena are the main problems for the reactor pressure vessel and the T-junctions both of them depend on the mixing of the coolant. The mixing process, flow and temperature distribution has been investigated experimentally using particle image velocimetry, laser induced fluorescence, and simulated by CFD tools. The obtained results showed that the ratio of flow rate between the main pipe and the branch pipe has a big influence on the mixing process. The particle image velocimetry/planar laser-induced fluorescence measurements technologies proved to be suitable for the investigation of turbulent mixing in the complicated flow system: both velocity and temperature distribution are important parameters in the determination of thermal fatigue and pressurized thermal shock. Results of the applied these techniques showed that both of them can be used as a good provider for data base and to validate CFD results.

  2. Effect on temperature of output of the core of the size of the break in the Upper Head of the vessel using TRACE5; Efecto sobre la Temperatura de Salida del Nucleo del Tamano de la Rotura en el Upper Head de la Vasija utilizando TRACE5

    Energy Technology Data Exchange (ETDEWEB)

    Querol, A.; Gallardo, S.; Verdu, G.

    2013-07-01

    Most (PWR) pressurized water reactors have thermocouples to detect overheating of the core since they are used to measure the temperature of exit of the nucleus (CET). However, it was found that in a small break (SBLOCA) located in the upper head of the vessel there is a delay between the measure of thermocouples and overheating of the core. This work is based on the simulation, using the code Thermo-hydraulic TRACE5, of the Test 6 - 1 the OECD/NEA rose project carried out in the experimental facility LSTF (Large Scale Test Facility). There have been different analyses in which geometric variables that can influence the model such as the size and location of the break, possible flow towards the break and the nodalization of the upper head of the vessel have been studied.

  3. Dicty_cDB: Contig-U01503-1 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available K*kkkk*kkkfkk*kn*lfiylkhdnikin *kyngk*kfnkqtne*fyk*fkl***fk*fy*wiy****tftnntwkfirnsdyvsk*mp lstf**ms*kilii*k...kl***fk*fy*wiy****tftnntwkfirnsdyvsk*mp lstf**ms*kilii*kkkkkylykkiiiiitiiinnnnniyvklklklklknk**kkkkk k Frame

  4. Commissioning of the ATLAS thermal-hydraulic integral test facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon-Sik [Thermal Hydraulics Safety Research Division, Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)], E-mail: yskim3@kaeri.re.kr; Choi, Ki-Yong; Park, Hyeon-Sik; Cho, Seok; Kim, Bok-Deug; Choi, Nam-Hyeon; Baek, Won-Pil [Thermal Hydraulics Safety Research Division, Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2008-10-15

    KAERI recently constructed a new thermal-hydraulic integral test facility for advanced pressurized water reactors (PWRs) - ATLAS. The ATLAS facility has the following characteristics: (a) 1/2-height and length, 1/288-volume, and full pressure simulation of APR1400, (b) maintaining a geometrical similarity with APR1400 including 2(hot legs) x 4(cold legs) reactor coolant loops, direct vessel injection (DVI) of emergency core cooling water, integrated annular downcomer, etc., (c) incorporation of specific design characteristics of OPR1000 such as cold leg injection and low-pressure safety injection pumps, (d) maximum 10% of the scaled nominal core power. The ATLAS will mainly be used to simulate various accident and transient scenarios for evolutionary PWRs, OPR1000 and APR1400: the simulation capability of broad scenarios including the reflood phase of a large-break loss-of-coolant accident (LOCA), small-break LOCA scenarios including DVI line breaks, a steam generator tube rupture, a main steam line break, a feed line break, a mid-loop operation, etc. The ATLAS is now in operation after an extensive series of commissioning tests in 2006.

  5. ORNL rod-bundle heat-transfer test data. Volume 6. Thermal-hydraulic test facility experimental data report for test 3. 05. 5B - double-ended cold-leg break simulation

    Energy Technology Data Exchange (ETDEWEB)

    Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.; Schwinkendorf, K.N.

    1982-05-18

    Thermal-Hydraulic Test Facility (THTF) Test 3.05.5B was conducted by members of the ORNL PWR Blowdown Heat Transfer Separate-Effects Program on July 3, 1980. The objective of the program is to investigate heat transfer phenomena believed to occur in PWRs during accidents, including small and large break loss-of-coolant accidents. Test 3.05.5B was designed to provide transient thermal-hydraulics data in rod bundle geometry under reactor accident-type conditions. Reduced instrument responses are presented. Also included are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers.

  6. Simulation with TRACE5 of a small break of 1% in the hot branch; Simulacion con TRACE5 de una Rotura Pequena del 1% en la Rama Caliente

    Energy Technology Data Exchange (ETDEWEB)

    Querol, A.; Gallardo, S.; Verdu, G.

    2013-07-01

    This work has been simulated, Thermo-hydraulic-coded TRACE5, the 1-2 Test of the OECD/NEA ROSA project that reproduces a break of 1% in the hot field of a water pressurized (PWR) reactor. The results are compared with the experimental values for studying the effect of the stratification of the liquid into hot branch, the geometry and the size of the break and flow injected by the HPI system.

  7. Post-test thermal-hydraulic analysis of two intermediate LOCA tests at the ROSA facility including uncertainty evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Freixa, J., E-mail: jordi@freixa.net [Paul Scherrer Institut (PSI) 5232 Villigen PSI (Switzerland); Kim, T.-W. [Paul Scherrer Institut (PSI) 5232 Villigen PSI (Switzerland); Manera, A. [University of Michigan, Ann Arbor, MI 48109 (United States)

    2013-11-15

    The OECD/NEA ROSA-2 project aims at addressing thermal-hydraulic safety issues relevant for light water reactors by building up an experimental database at the ROSA Large Scale Test Facility (LSTF). The ROSA facility simulates a PWR Westinghouse design with a four-loop configuration and a nominal power of 3423 MWth. Two intermediate break loss-of-coolant-accident (LOCA) experiments (Tests 1 and 2) have been carried out during 2010. The two tests were analyzed by using the US-NRC TRACE best estimate code, employing the same nodalization previously used for the simulation of small-break LOCA experiments of the ROSA-1 programme. A post-test calculation was performed for each test along with uncertainty analysis providing uncertainty bands for each relevant time trend. Uncertainties in the code modelling capabilities as well as in the initial and boundary conditions were taken into account, following the guidelines and lessons learnt through participation in the OECD/NEA BEMUSE programme. Two different versions of the TRACE code were used in the analysis, providing a qualitatively good prediction of the tests. However, the uncertainty analysis revealed differences between the performances of some models in the two versions. The most relevant parameters of the two experimental tests were falling within the computed uncertainty bands.

  8. Post-test thermal-hydraulic analysis of two intermediate LOCA tests at the ROSA facility including uncertainty evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Freixa, J.; Kim, T-W.; Manera, A. [Paul Scherrer Inst., Villigen (Switzerland)

    2011-07-01

    The OECD/NEA ROSA-2 project aims at addressing thermal-hydraulic safety issues relevant for light water reactors by building up an experimental database at the ROSA Large Scale Test Facility (LSTF). The ROSA facility simulates a PWR Westinghouse design with a four-loop configuration and a nominal power of 3423 MWth. Two intermediate break loss-of-coolant-accident (LOCA) experiments (Test 1 and 2) have been carried out during 2010. The two tests were analyzed by using the US-NRC TRACE best estimate code, employing the same nodalization previously used for the simulation of small-break LOCA experiments of the ROSA-1 program. A post-test calculation was performed for each test along with uncertainty analysis providing uncertainty bands for each relevant time trend. Uncertainties in the code modeling capabilities as well as in the initial and boundary conditions were taken into account, following the guidelines and lessons learnt through participation in the OECD/NEA BEMUSE program. Two different versions of the TRACE code were used in the analysis, providing a qualitatively good prediction of the tests. However, both versions showed deficiencies that need to be addressed. The most relevant parameters of the two experimental tests were falling within the computed uncertainty bands. (author)

  9. SECOND ATLAS DOMESTIC STANDARD PROBLEM (DSP-02 FOR A CODE ASSESSMENT

    Directory of Open Access Journals (Sweden)

    YEON-SIK KIM

    2013-12-01

    Full Text Available KAERI (Korea Atomic Energy Research Institute has been operating an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS, for transient and accident simulations of advanced pressurized water reactors (PWRs. Using ATLAS, a high-quality integral effect test database has been established for major design basis accidents of the APR1400 plant. A Domestic Standard Problem (DSP exercise using the ATLAS database was promoted to transfer the database to domestic nuclear industries and contribute to improving a safety analysis methodology for PWRs. This 2nd ATLAS DSP (DSP-02 exercise aims at an effective utilization of an integral effect database obtained from ATLAS, the establishment of a cooperation framework among the domestic nuclear industry, a better understanding of the thermal hydraulic phenomena, and an investigation into the possible limitation of the existing best-estimate safety analysis codes. A small break loss of coolant accident with a 6-inch break at the cold leg was determined as a target scenario by considering its technical importance and by incorporating interests from participants. This DSP exercise was performed in an open calculation environment where the integral effect test data was open to participants prior to the code calculations. This paper includes major information of the DSP-02 exercise as well as comparison results between the calculations and the experimental data.

  10. Joint research project WASA-BOSS: Further development and application of severe accident codes. Assessment and optimization of accident management measures. Project B: Accident analyses for pressurized water reactors with the application of the ATHLET-CD code; Verbundprojekt WASA-BOSS: Weiterentwicklung und Anwendung von Severe Accident Codes. Bewertung und Optimierung von Stoerfallmassnahmen. Teilprojekt B: Druckwasserreaktor-Stoerfallanalysen unter Verwendung des Severe-Accident-Codes ATHLET-CD

    Energy Technology Data Exchange (ETDEWEB)

    Jobst, Matthias; Kliem, Soeren; Kozmenkov, Yaroslav; Wilhelm, Polina

    2017-02-15

    Within the framework of the project an ATHLET-CD input deck for a generic German PWR of type KONVOI has been created. This input deck was applied to the simulation of severe accidents from the accident categories station blackout (SBO) and small-break loss-of-coolant accidents (SBLOCA). The complete accident transient from initial event at full power until the damage of reactor pressure vessel (RPV) is covered and all relevant severe accident phenomena are modelled: start of core heat up, fission product release, melting of fuel and absorber material, oxidation and release of hydrogen, relocation of molten material inside the core, relocation to the lower plenum, damage and failure of the RPV. The model has been applied to the analysis of preventive and mitigative accident management measures for SBO and SBLOCA transients. Therefore, the measures primary side depressurization (PSD), injection to the primary circuit by mobile pumps and for SBLOCA the delayed injection by the cold leg hydro-accumulators have been investigated and the assumptions and start criteria of these measures have been varied. The time evolutions of the transients and time margins for the initiation of additional measures have been assessed. An uncertainty and sensitivity study has been performed for the early phase of one SBO scenario with PSD (until the start of core melt). In addition to that, a code -to-code comparison between ATHLET-CD and the severe accident code MELCOR has been carried out.

  11. Simulation of a beyond design-basis-accident with RELAP5/MOD3.1

    Energy Technology Data Exchange (ETDEWEB)

    Banati, J. [Lappeenranta Univ. of Technology, Lappeenranta (Finland)

    1995-09-01

    This paper summarizes the results of analyses, parametric and sensitivity studies, performed using the RELAP5/MOD3.1 computer code for the 4th IAEA Standard Problem Exercise (SPE-4). The test, conducted on the PMK-2 facility in Budapest, involved simulation of a Small Break Loss Of Coolant Accident (SBLOCA) with a 7.4% break in the cold leg of a VVER-440 type pressurized water reactor. According to the scenario, the unavailability of the high pressure injection system led to a beyond design basis accident. For prevention of core damage, secondary side bleed-and-feed accident management measures were applied. A brief description of the PMK-2 integral type test facility is presented, together with the profile and some key phenomenological aspects of this particular experiment. Emphasis is placed on the ability of the code to predict the main trends observed in the test and thus, an assessment is given for the code capabilities to represent the system transient.

  12. Pre-test analysis of an integral effect test facility for thermal-hydraulic similarities of 6 inches coldleg break and DVI injection line break using MARS-1D

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Tae Soon; Choi, Ki Yong; Park, Hyun Sik; Euh, Dong Jin; Baek, Won Pil [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-03-01

    A pre-test analysis of a small-break loss-of-coolant accident (SBLOCA, DVI Line break) has been performed for the integral effect test loop of Korea Atomic Energy Research Institute (Korea Atomic Energy Research Institute-ITL), the construction of which will be started soon. The Korea Atomic Energy Research Institute-ITL is a full-height and 1/310 volume-scaled test facility based on the design features of the APR1400 (Korean Next Generation Reactor). This paper briefly introduces the basic design features of the Korea Atomic Energy Research Institute-ITL and presents the results of pre-test analysis for a postulated cold leg SBLOCA and DVI line break. Based on the same control logics and accident scenarios, the similarity between the Korea Atomic Energy Research Institute-ITL and the prototype plant, APR1400, is evaluated by using the MARS code, which is a multi-dimensional best-estimate thermal hydraulic code being developed by Korea Atomic Energy Research Institute. It is found that the Korea Atomic Energy Research Institute-ITL and APR 1400 have similar thermal hydraulic responses against the analyzed SBLOCA and DVI Line break scenario. It is also verified that the volume scaling law, applied to the design of the Korea Atomic Energy Research Institute-ITL, gives a reasonable results to keep a similarity with APR1400. 11 refs., 19 figs., 3 tabs. (Author)

  13. Fabrication of a Microtubular La0.6Sr0.4Ti0.2Fe0.8O3−δ Membrane by Electrophoretic Deposition for Hydrogen Production

    Directory of Open Access Journals (Sweden)

    Kyoung-Jin Lee

    2015-01-01

    Full Text Available Microtubular type La0.6Sr0.4Ti0.2Fe0.8O3−δ (LSTF membranes were prepared by electrophoretic deposition (EPD. The oxygen permeation and hydrogen production behavior of the membranes were investigated under various conditions. LSTF green layer was successfully coated onto a carbon rod and, after heat treatment at 1400°C in air, a dense LSTF tubular membrane with a thickness of 250 mm can be obtained. The oxygen permeation and hydrogen production rate were enhanced by CH4 in the permeate side, and the hydrogen production rate by water splitting was 0.22 mL/min·cm2 at 1000°C. It is believed that hydrogen production via water splitting using these tubular LSTF membranes is possible.

  14. Assessment of passive safety injection systems of ALWRs. Final report of the European Commission 4th framework programme. Project FI4I-CT95-004 (APSI)

    Energy Technology Data Exchange (ETDEWEB)

    Tuunanen, J. [VTT Energy, Espoo (Finland). Nuclear Energy; Vihavainen, J. [Lappeenranta Univ. of Technology (Finland); D' Auria, F. [Univ. of Pisa (Italy); Kimber, G. [AEA Technology (United Kingdom)

    1999-07-01

    The European Commission 4th Framework Programme project 'Assessment of Passive Safety Injection Systems of Advanced Light Water Reactors (FI4I-CT95-0004)' involved experiments on the PACTEL test facility and computer simulations of selected experiments. The experiments focused on the performance of Passive Safety Injection Systems (PSIS) of Advanced Light Water Reactors (ALWRs) in Small Break Loss-Of-Coolant Accident (SBLOCA) conditions. The PSIS consisted of a Core Make-up Tank (CMT) and two pipelines. A pressure balancing line (PBL) connected the CMT to one cold leg. The injection line (IL) connected it to the downcomer. The project involved 15 experiments in three series. The experiments provided valuable information about condensation and heat transfer processes in the CMT, thermal stratification of water in the CMT, and natural circulation flow through the PSIS lines. The experiments showed the examined PSIS works efficiently in SBLOCAs although the flow through the PSIS may stop in very small SBLOCAs, when the hot water fills the CMT. The experiments also demonstrated the importance of flow distributor (sparger) in the CMT to limit rapid condensation. The project included validation of three thermal-hydraulic computer codes (APROS, CATHARE and RELAP5). The analyses showed the codes are capable of simulating the overall behaviour of the transients. The codes predicted accurately the core heatup, which occurred when the primary coolant inventory was reduced so much that the core top became free of water. The detailed analyses of the calculation results showed that some models in the codes still need improvements. Especially, further development of models for thermal stratification, condensation and natural circulation flow with small driving forces would be necessary for accurate simulation of phenomena in the PSIS. (orig.)

  15. Comparison report of open calculations for ATLAS Domestic Standard Problem (DSP 02)

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Ki Yong; Kim, Y. S.; Kang, K. H.; Cho, S.; Park, H. S.; Choi, N. H.; Kim, B. D.; Min, K. H.; Park, J. K.; Chun, H. G.; Yu, Xin Guo; Kim, H. T.; Song, C. H.; Sim, S. K.; Jeon, S. S.; Kim, S. Y.; Kang, D. G.; Choi, T. S.; Kim, Y. M.; Lim, S. G.; Kim, H. S.; Kang, D. H.; Lee, G. H.; Jang, M. J.

    2012-09-15

    KAERI (Korea Atomic Energy Research Institute) has been operating an integral effect test facility, the Advanced Thermal Hydraulic Test Loop for Accident Simulation (ATLAS) for transient and accident simulations of advanced pressurized water reactors (PWRs). By using the ATLAS, a high quality integral effect test database has been established for major design basis accidents of the APR1400. A Domestic Standard Problem (DSP) exercise using the ATLAS database was promoted in order to transfer the database to domestic nuclear industries and to contribute to improving safety analysis methodology for PWRs. This 2nd ATLAS DSP exercise was led by KAERI in collaboration with KINS since the successful completion of the 1st ATLAS DSP in 2009. This exercise aims at effective utilization of integral effect database obtained from the ATLAS, establishment of cooperation framework among the domestic nuclear industry, better understanding of thermal hydraulic phenomena, and investigation of the possible limitation of the existing best estimate safety analysis codes. A small break loss of coolant accident of 6 inch break at the cold leg was determined as a target scenario by considering its technical importance and by incorporating with interests from participants. Twelve domestic organizations joined this DSP 02 exercise. Finally, eleven out of the joined organizations submitted their calculation results, including universities, government, and nuclear industries. This DSP exercise was performed in an open calculation environment where the integral effect test data was open to participants prior to code calculations. This report includes all information of the 2nd ATLAS DSP (DSP 02) exercise as well as comparison results between the calculations and the experimental data.

  16. SPES-2, AP600 intergral system test S01007 2 inch CL to core make-up tank pressure balance line break

    Energy Technology Data Exchange (ETDEWEB)

    Bacchiani, M.; Medich, C.; Rigamonti, M. [SIET S.p.A. Piacenza (Italy)] [and others

    1995-09-01

    The SPES-2 is a full height, full pressure experimental test facility reproducing the Westinghouse AP600 reactor with a scaling factor of 1/395. The experimental plant, designed and operated by SIET in Piacenza, consists of a full simulation of the AP600 primary core cooling system including all the passive and active safety systems. In 1992, Westinghouse, in cooperation with ENEL (Ente Nazionale per l` Energia Elettrica), ENEA (Enter per le numove Technlogie, l` Energia e l` Ambient), Siet (Societa Informazioni Esperienze Termoidraulich) and ANSALDO developed an experimental program to test the integrated behaviour of the AP600 passive safety systems. The SPES-2 test matrix, concluded in November 1994, has examined the AP600 passive safety system response for a range of small break LOCAs at different locations on the primary system and on the passive system lines; single steam generator tube ruptures with passive and active safety systems and a main steam line break transient to demonstrate the boration capability of passive safety systems for rapid cooldown. Each of the tests has provided detailed experimental results for verification of the capability of the analysis methods to predict the integrated passive safety system behaviour. Cold and hot shakedown tests have been performed on the facility to check the characteristics of the plant before starting the experimental campaign. The paper first presents a description of the SPES-2 test facility then the main results of S01007 test {open_quotes}2{close_quotes} Cold Leg (CL) to Core Make-up Tank (CMT) pressure balance line break{close_quotes} are reported and compared with predictions performed using RELAP5/mod3/80 obtained by ANSALDO through agreement with U.S.N.R.C. (U.S. Nuclear Regulatory Commission). The SPES-2 nodalization and all the calculations here presented were performed by ANSALDO and sponsored by ENEL as a part of pre-test predictions for SPES-2.

  17. A study of return to saturation oscillations in the OSU APEX thermal hydraulic testing facility

    Science.gov (United States)

    Franz, Scott Cameron

    The purpose of this paper is to describe the flow oscillations which occur in the AP600 long term cooling test facility at Oregon State University. The AP600 system is an advanced pressurized water reactor design utilizing passive emergency cooling systems. A few hours after the initiation of a cold leg break, the passive cooling systems inject gravity fed cold water at a rate allowing steam production in the reactor vessel. Steam production in the core causes the pressure in the upper head to increase leading to flow oscillations in all the connecting reactor systems. This paper will show that the oscillations have a definite region of onset and termination for specific conditions in the APEX testing facility. Tests performed at high powers, high elevation breaks, and small break sizes do not exhibit oscillations. The APOS (Advanced Plant Oscillation Simulator) computer code has been developed using a quasi-steady state analysis for flows and a transient analysis for the core node energy balance. The pressure in the reactor head is calculated using a modified perfect gas analysis. For tank liquid inventories, a simple conservation of mass analysis is used to estimate the tank elevations. Simulation logic gleaned from APEX data and photographic evidence have been incorporated into the code to predict termination of the oscillations. Areas which would make the work more complete include a better understanding of two-phase fluid behavior for a top offtake on a pipe, more instrumentation in the core region of the APEX testing facility, and a clearer understanding of fluid conditions in the reactor barrel. Scaling of the oscillations onset and pressure amplitude are relatively straightforward, but termination and period are difficult to scale to the full AP600 plant. Differences in the core power profile and other geometrical differences between the testing facility and the actual plant make the scaling of this phenomenon to the actual plant conditions very difficult.

  18. LMFBR with booster pump in pumping loop

    Science.gov (United States)

    Rubinstein, H.J.

    1975-10-14

    A loop coolant circulation system is described for a liquid metal fast breeder reactor (LMFBR) utilizing a low head, high specific speed booster pump in the hot leg of the coolant loop with the main pump located in the cold leg of the loop, thereby providing the advantages of operating the main pump in the hot leg with the reliability of cold leg pump operation.

  19. Thermal and fluid mixing in a 1/2-scale test facility. Volume 1. Facility and test design report. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Dolan, F.X.; Valenzuela, J.A.

    1985-07-01

    This report describes the test facility and program designed to measure fluid mixing and heat transfer in a 1/2-scale model of the cold leg downcomer and lower plenum of a pressurized water reactor under conditions of interest to the issue of pressurized thermal shock. Several cold leg assemblies are modeled and the downcomer arrangement can be altered to match vendor-specific configurations. The facility can be operated to model flow rates based on Froude number of the injected flow in the cold leg and with steady or transient inlet boundary conditions. Extensive instrumentation is provided to measure flow rates, temperatures and pressure at the facility boundaries and for detailed measurements of temperature, velocity and heat transfer data in the cold leg and downcomer models. The test data are monitored and recorded by a computer data acquisition system that is also used for post-test data reduction and plotting. The planned test matrix includes 75 tests with variations in cold leg and downcomer geometries, loop and HPI flow rates, cold leg Froude number and loop to HPI density difference. Test results will be reported in a series of quick look and final report.

  20. TRAC PF1/MOD1 calculations and data comparisons for mist feed and bleed and steam generator tube rupture experiments

    Energy Technology Data Exchange (ETDEWEB)

    Siebe, D.A.; Boyack, B.E.; Steiner, J.L.

    1988-01-01

    Los Alamos National Laboratory is a participant in the Integral System Test (IST) program initiated in June 1983 for the purpose of providing integral system test data on specific issues/phenomena relevant to post-small-break loss-of-coolant accidents, loss of feedwater and other transients in Babcock and Wilcox (BandW) plant designs. The Multi-Loop Integral System Test (MIST) facility is the largest single component in the IST program. MIST is a 2 /times/ 4 (two hot legs and steam generators (SGs), four cold legs and reactor coolant pumps) representation of lowered-loop reactor system of the BandW design. It is a full-height, full-pressure facility with 1/817 power and volume scaling. Two other integral experimental facilities are included in the IST program: test loops at the University of Maryland, College Park, and at SRI International (SRI-2). The objective of the IST tests is to generate high-quality experimental data to be used for assessing thermal-hydraulic safety computer codes. Efforts are under way at Los Alamos to assess TRAC-PF1/MOD1 against data from each of the IST facilities. Calculations and data comparisons for TRAC-PF1/MOD1 assessment are presented for two transients run in the MIST facility. These are MIST Test 330302, a feed and bleed test with delayed high-pressure injection; and Test 3404AA, an SG tube-rupture test with the affected SG isolated. Only MIST assessment results are presented in this paper. The TRAC-PF1/MOD1 calculations completed to date for MIST tests are in reasonable agreement with the data from these tests. Reasonable agreement is defined as meaning that major trends are predicted correctly, although TRAC values are frequently outside the range of data uncertainty. We believe that correct conclusions will be reached if the code is used in similar applications despite minor code/model deficiencies. 7 refs., 5 figs., 2 tabs.

  1. 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

  2. Water Reflooding Effectiveness Assessment for 1 000 MWe PWR under Severe Accident Condition%百万千瓦级压水堆严重事故后再注水的有效性评价

    Institute of Scientific and Technical Information of China (English)

    胡啸; 黄挺; 裴杰; 陈炼

    2015-01-01

    根据现有的设计资料,使用一体化严重事故分析程序 MELCOR1.8.6建立了核电厂一、二回路系统,非能动堆芯冷却系统和安全壳系统的模型,并模拟冷段2英寸(5.08 cm)小破口叠加重力注入失效的严重事故发生后,将冷却剂注入堆芯的情形,分析其对严重事故进程的缓解能力。本文选取3个严重事故的不同阶段,将冷却剂分别以小流量(10 kg/s)、中流量(50 kg/s)和大流量(200 kg/s)的速率注入堆芯,通过比较氢气产生量、堆芯放射性产生量及堆芯温度等数据来评估在严重事故不同阶段再注水的可行性。结果表明:在堆芯损伤初期,可认为10 kg/s以上的流量足以冷却百万千瓦级事故安全。而当严重事故发展到堆芯开始坍塌阶段,200 kg/s的注水流量可认为是基本可行的,而小于此流量的注水应慎重考虑。%The MELCOR1.8.6 code was applied to a severe accident model of a 1 000 MWe PWR which includes primary system,secondary system,passive core cool-ing system and containment system.For the transient case,a small break LOCA with 2 inch (5.08 cm)break at the cold leg concurrent with failure of gravity injection was selected.After the core was damaged due to the failure of gravity inj ection,it was assumed that the coolant was inj ected into the pressure vessel,and then the water reflooding effectiveness was evaluated and analyzed.In this calculation,the coolant injection into reactor core with the small (10 kg/s),medium (50 kg/s)and large (200 kg/s)mass flow rates respectively at 3 different time stages of the severe accident was simulated.The effectiveness of water reflooding was assessed through hydrogen production,radioactive materials released from core,and core temperature.The results show that the mass flow rate above 10 kg/s is believed to be efficient for cooling a 1 000 MWe reactor at the beginning of core damage.However,with the accident devel-oping to core relocation,a large mass flow

  3. Experimental data report for transient flow calibration facility tests IA101, AI102 and IA103

    Science.gov (United States)

    Martinell, J. S.; Wambach, J. L.; Crapo, H. S.

    1980-03-01

    Thermal hydraulic response data are presented for the transient performance tests of a modular Drag Disc - Turbine transducer rake. The tests were conducted in a system which provided full scale simulation of the pressure vessel and broken loop cold leg piping of the Loss of Fluid Test Facility. A low cell system was used to provide a reference mass flow rate measurement.

  4. Studies on Perovskite-Based Electrodes for Symmetrical SOFCs

    Directory of Open Access Journals (Sweden)

    Dos Santos García, A. J.

    2008-10-01

    Full Text Available The use of the same material as anode and cathode in symmetrical solid oxide fuel cells (SFCs promises notable benefits as easier fabrication, hence lower cost production and resistance to carbon formation upon fuel cracking. Although chromites and chromo-manganites have been proposed as candidate electrode materials for this novel SOFC configuration, demonstrating promising performances, further work is required to develop compositions exhibiting higher efficiencies. In the present work we evaluate the structural evolution from cubic to orthorhombic unit cells with increasing the Fe content and the performance of La4Sr8Ti12-xFexO38-δ (LSTF phases and compare their response with other symmetrical electrodes. The electrochemical performance is 20% higher when using graded LSTF electrodes than in other perovskite-based systems.

    La utilización simultánea de un mismo material cerámico como ánodo y cátodo en pilas de combustible de óxido sólido simétricas (SFCs aporta una serie de beneficios entre los que figura una fabricación más sencilla, reducción de los costes de producción, así como resistencia a la formación de depósitos de carbón por craqueo del combustible. Recientemente, cromitas y cromomanganitas han sido propuestos como materiales capaces de adoptar esta novedosa configuración SOFC y, si bien los resultados obtenidos son prometedores, se requiere de una mayor investigación para el desarrollo de nuevas composiciones que presenten eficiencias más elevadas. En el presente trabajo, se evalúan la evolución de la estructura desde celdas cúbicas a ortorrómbicas al aumentar el contenido en Fe y las prestaciones del sistema La4Sr8Ti12-xFexO38-δ (LSTF y se compara su respuesta con otros electrodos simétricos, observándose que el rendimiento es hasta un 20% mayor en el caso de emplear electrodos LSTF que en

  5. An experience with in-service fabrication and inspection of austenitic stainless steel piping in high temperature sodium system

    Energy Technology Data Exchange (ETDEWEB)

    Ravi, S., E-mail: sravi@igcar.gov.in; Laha, K.; Sakthy, S.; Mathew, M.D.; Bhaduri, A.K.

    2015-04-01

    Highlights: • Procedure for changing 304L SS pipe to 316L SS in sodium loop has been established. • Hot leg made of 304L SS was isolated from existing cold leg made of 316LN SS. • Innovative welding was used in joining the new 316L SS pipe with existing 316LN SS. • The old components of 304L SS piping have been integrated with the new piping. - Abstract: A creep testing facility along with dynamic sodium loop was installed at Indira Gandhi Centre for Atomic Research, Kalpakkam, India to assess the creep behavior of fast reactor structural materials in flowing sodium. Type 304L austenitic stainless steel was used in the low cross section piping of hot-leg whereas 316LN austenitic stainless steel in the high cross section cold-leg of the sodium loop. The intended service life of the sodium loop was 10 years. The loop has performed successfully in the stipulated time period. To enhance its life time, it has been decided to replace the 304L piping with 316L piping in the hot-leg. There were more than 300 welding joints involved in the integration of cold-leg with the new 316L hot-leg. Continuous argon gas flow was maintained in the loop during welding to avoid contamination of sodium residue with air. Several innovative welding procedures have been adopted for joining the new hot-leg with the existing cold-leg in the presence of sodium residue adopting TIG welding technique. The joints were inspected for 100% X-ray radiography and qualified by performing tensile tests. The components used in the discarded hot-leg were retrieved, cleaned and integrated in the renovated loop. A method of cleaning component of sodium residue has been established. This paper highlights the in-service fabrication and inspection of the renovation.

  6. Error detection in core loading in the condition of asymmetrical distribution of power

    Energy Technology Data Exchange (ETDEWEB)

    Ryzhov, Andrei; Pinegin, Anatoly [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)

    2016-09-15

    The error detection in core loading in many cases happens in conditions of a significant asymmetry in the distribution of power density, which is caused by thermal mechanical deformations of reactor core, and temperature differences in the cold legs of coolant. The asymmetry of power distribution essentially decreases the effectiveness of algorithms used to detect errors in the core loading using ICIS data. The paper proposes the ways for solving this problem by means of special filtration algorithms.

  7. Simulation with RELAP5/MOD3.3 of a postulated 10% hot leg break in Angra 2 nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Azevedo, Carlos Vicente Goulart de; Palmieri, Elcio Tadeu; Aronne, Ivan Dionysio [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)], e-mail: cvga@cdtn.br, e-mail: etp@cdtn.br, e-mail: aroneid@cdtn.br

    2009-07-01

    This paper presents the simulation results of a 10% break in the hot leg of Angra 2 nuclear power plant, which was run with the computer code RELAP5/MOD3.3. The initial steady state conditions for this simulation are in agreement with the experiment named SB-HL-02 that was conducted in the Large Scale Test Facility in the Rig of Safety Assessment-IV program (ROSA-IV/LSTF). The main boundary conditions specified for the simulation were: high pressure injection system (HPI) and auxiliary feedwater system (AFW) were assumed to be unavailable; and loss of offsite power was assumed to occur concurrently with scram. The results obtained were scaled down and compared with the ROSA-IV/LSTF test, which was performed with the same boundary conditions. This activity was executed in the scope of IAEA research project (CRP J72005) - Evaluation of Uncertainties in the Simulation of Accidents in Angra 2 using RELAP5/MOD3.3 Code Applying CIAU Methodology. (author)

  8. Measuring and modeling suspended sediment concentration profiles in the surf zone

    Institute of Scientific and Technical Information of China (English)

    2012-01-01

    Time-averaged suspended sediment concentration profiles across the surf zone were measured in a large-scale three-dimensional movable bed laboratory facility (LSTF:Large-scale Sediment Transport Facility). Sediment suspension under two different types of breaking waves, spilling and plunging breakers, was investigated. The magnitudes and shapes of the concentration profiles varied substantially at different locations across the surf zone, reflecting the different intensities of breaking-induced turbulence. Sediment sus- pension at the energetic plunging breaker-line was much more active, resulting in nearly homogeneous concentration profiles throughout most of the water column, as compared to the reminder of the surf zone and at the spilling breaker-line. Four suspended sediment concentration models were examined based on the LSTF data, including the mixing turbulence length approach, segment eddy viscosity model, breaking-induced wave-energy dissipation approach, and a combined breaking and turbulence length model developed by this study. Neglecting the breaking-induced turbulence and subsequent sediment mixing, suspended sediment concentration models failed to predict the across-shore variations of the sediment suspension, especially at the plunging breaker-line. Wave-energy dissipation rate provided an accurate method for estimating the intensity of turbulence generated by wave breaking. By incorporating the breaking-induced turbulence, the combined breaking and turbulence length model reproduced the across-shore variation of sediment suspension in the surf zone. The combined model reproduced the measured time-averaged suspended sediment concentration profiles reasonably well across the surf zone.

  9. Simulation of a transient with loss of primary coolant due to a small rupture in Angra 2 nuclear power plant with RELAP5/MOD3.2.2G code; Simulacao de um acidente postulado de perda de refrigerante primario por pequena ruptura na usina de Angra 2 com o codigo RELAP5/MOD3.2.2G

    Energy Technology Data Exchange (ETDEWEB)

    Sabundjian, Gaiane; Andrade, Delvonei Alves de [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

    2002-07-01

    This paper presents a nodalization for Angra 2 Nuclear Power Plant, as well as the results obtained for a Small Break Loss of Coolant Accident (SBLOCA), simulated with RELAP5/MOD3.2G code. This accident consists in a small break (380 m{sup 2}) in the line of the Emergency Core Coolant System (ECCS) in loop 20 of Angra 2. Results are not as expected, however they are satisfactory regarding the nodalization used. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Morlent, O.; Reuchet, J. [CEA Fontenay-aux-Roses, Inst. de Protection et de Surete Nucleaire, 92 (France)

    2001-07-01

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

  11. Two-phase CFD PTS validation in an extended range of thermohydraulics conditions covered by the COSI experiment

    Energy Technology Data Exchange (ETDEWEB)

    Coste, P., E-mail: pierre.coste@cea.fr [CEA, DEN/DANS/DM2S/SMTF/LMSF, 17 rue des Martyrs, 38054 Grenoble (France); Ortolan, A. [CEA, DEN/DANS/DM2S/SMTF/LMSF, 17 rue des Martyrs, 38054 Grenoble (France); ENSICA Engineering School, Toulouse (France)

    2014-11-15

    Highlights: • Models for large interfaces in two-phase CFD were developed for PTS. • The COSI experiment is used for NEPTUNE{sub C}FD integral validation. • COSI is a PWR cold leg scaled 1/100 for volume. • Fifty runs are calculated, covering a large range of flow configurations. • The CFD predicting capability is analysed using global and local measurements. - Abstract: In the context of the Pressurized Water Reactors (PWR) life duration safety studies, some models were developed to address the Pressurized Thermal Shock (PTS) from the two-phase CFD angle, dealing with interfaces much larger than cells size and with direct contact condensation. Such models were implemented in NEPTUNE{sub C}FD, a 3D transient Eulerian two-fluid model. The COSI experiment is used for its integral validation. It represents a cold leg scaled 1/100 for volume and power from a 900 MW PWR under a large range of LOCA PTS conditions. In this study, the CFD is evaluated in the whole range of parameters and flow configurations covered by the experiment. In a first step, a single choice of mesh and CFD models parameters is fixed and justified. In a second step, fifty runs are calculated. The CFD predicting capability is analysed, comparing the liquid temperature and the total condensation rate with the experiment, discussing their dependency on the inlet cold liquid rate, on the liquid level in the cold leg and on the difference between co-current and counter-current runs. It is shown that NEPTUNE{sub C}FD 1.0.8 calculates with a fair agreement a large range of flow configurations related to ECCS injection and steam condensation.

  12. Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Papp, L. [Inst. of Material Engineering, Ostrava (Switzerland)

    1995-12-31

    After modification of Dukovany NPP steam generator (SG) feedwater system, the increased concentration of minerals was measured in the cold leg of modified SG. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators has focused the attention to the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of the flow distribution in the secondary side of SG was developed.

  13. Statistical analysis of the blowdown phase of a loss-of-coolant accident in a pressurized water reactor as calculated by RELAP4/MOD6

    Energy Technology Data Exchange (ETDEWEB)

    Berman, M.; Byers, R.K.; Steck, G.P.

    1979-01-01

    A statistical study is presented of the blowdown phase of a design basis accident (double-ended cold leg guillotine break) in the Zion pressurized water reactor. The response surface method was employed to generate a polynomial approximation of the peak clad temperatures calculated by RELAP4/MOD6. The nodalization was a modification of the RELAP model of Zion developed in the BE/EM study. Twenty one variables were initially selected for the study. These variables, their ranges and distributions resulted from the best engineering judgement of NRC, Sandia, INFL, and other interested and knowledgeable investigators.

  14. Evaluation of the thermal-hydraulic response and fuel rod thermal and mechanical deformation behavior during the power burst facility test LOC-3. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Yackle, T.R.; MacDonald, P.E.; Broughton, J.M.

    1980-01-01

    An evaluation of the results from the LOC-3 nuclear blowdown test conducted in the Power Burst Facility is presented. The test objective was to examine fuel and cladding behavior during a postulated cold leg break accident in a pressurized water reactor (PWR). Separate effects of rod internal pressure and the degree of irradiation were investigated in the four-rod test. Extensive cladding deformation (ballooning) and failure occurred during blowdown. The deformation of the low and high pressure rods was similar; however, the previously irradiated test rod deformed to a greater extent than a similar fresh rod exposed to identical system conditions.

  15. Large scale breeder reactor plant prototype mechanical pump conceptual design study

    Energy Technology Data Exchange (ETDEWEB)

    1976-07-01

    This final report is a complete conceptual design study of a mechanical pump for a large scale breeder reactor plant. The pumps are located in the cold leg side of the loops. This makes the net positive suction head available - NPSHA - low, and is, in fact, a major influencing factor in the design. Where possible, experience gained from the Clinch River Project and the FFTF is used in this study. Experience gained in the design, manufacturer, and testing of pumps in general and sodium pumps in particular is reflected in this report. The report includes estimated cost and time schedule for design, manufacture, and testing. It also includes a recommendation for development needs.

  16. Recriticality of reactor during cooling post loss of coolant accident due to great ruptures; Recriticalidade do reator durante resfriamento apos um acidente de perda de refrigeracao por grandes rupturas

    Energy Technology Data Exchange (ETDEWEB)

    Ponzoni Filho, Pedro [Universidade Federal, Rio de Janeiro, RJ (Brazil). Fundacao Universitaria Jose Bonifacio

    2000-07-01

    The reduction of boron concentration in Angra 1 Boron Injection Tank (BIT) to the room temperature solubility level makes necessary a recalculation, to meet the post-LOCA subcriticality requirement. The main conclusions are the following: the new boron concentration of accumulators, Refueling Water Storage Tank (RWST) and BIT must be 2500 to 3000 ppm; all the devices used for heating BIT and the lines and valves associated to it may be turned off; the time for switch over from cold leg to hot leg recirculation must change from 24 to 13 hours; the new range of boron concentration will assure subcriticality of Angra 1 during a LOCA long term reactor cooling. (author)

  17. Improvements in the simulation of a main steam line break with steam generator tube rupture

    Science.gov (United States)

    Gallardo, Sergio; Querol, Andrea; Verdú, Gumersindo

    2014-06-01

    The result of simultaneous Main Steam Line Break (MSLB) and a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR) is a depressurization in the secondary and primary system because both systems are connected through the SGTR. The OECD/NEA ROSA-2 Test 5 performed in the Large Scale Test Facility (LSTF) reproduces these simultaneous breaks in a Pressurized Water Reactor (PWR). A simulation of this Test 5 was made with the thermal-hydraulic code TRACE5. Some discrepancies found, such as an underestimation of SG-A secondary pressure during the depressurization and overestimation of the primary pressure drop after the first Power Operated Relief Valve (PORV) opening can be improved increasing the nodalization of the Upper Head in the pressure vessel and meeting the actual fluid conditions of Upper Head during the transient.

  18. Measuring and modeling suspended sediment concentration profiles in the surf zone

    Directory of Open Access Journals (Sweden)

    Ping Wang

    2012-10-01

    Full Text Available Time-averaged suspended sediment concentration profiles across the surf zone were measured in a large-scale three-dimensional movable bed laboratory facility (LSTF: Large-scale Sediment Transport Facility. Sediment suspension under two different types of breaking waves, spilling and plunging breakers, was investigated. The magnitudes and shapes of the concentration profiles varied substantially at different locations across the surf zone, reflecting the different intensities of breaking-induced turbulence. Sediment suspension at the energetic plunging breaker-line was much more active, resulting in nearly homogeneous concentration profiles throughout most of the water column, as compared to the reminder of the surf zone and at the spilling breaker-line. Four suspended sediment concentration models were examined based on the LSTF data, including the mixing turbulence length approach, segment eddy viscosity model, breaking-induced wave-energy dissipation approach, and a combined breaking and turbulence length model developed by this study. Neglecting the breaking-induced turbulence and subsequent sediment mixing, suspended sediment concentration models failed to predict the across-shore variations of the sediment suspension, especially at the plunging breaker-line. Wave-energy dissipation rate provided an accurate method for estimating the intensity of turbulence generated by wave breaking. By incorporating the breaking-induced turbulence, the combined breaking and turbulence length model reproduced the across-shore variation of sediment suspension in the surf zone. The combined model reproduced the measured time-averaged suspended sediment concentration profiles reasonably well across the surf zone.

  19. CFD Code Validation against Stratified Air-Water Flow Experimental Data

    Directory of Open Access Journals (Sweden)

    F. Terzuoli

    2008-01-01

    Full Text Available Pressurized thermal shock (PTS modelling has been identified as one of the most important industrial needs related to nuclear reactor safety. A severe PTS scenario limiting the reactor pressure vessel (RPV lifetime is the cold water emergency core cooling (ECC injection into the cold leg during a loss of coolant accident (LOCA. Since it represents a big challenge for numerical simulations, this scenario was selected within the European Platform for Nuclear Reactor Simulations (NURESIM Integrated Project as a reference two-phase problem for computational fluid dynamics (CFDs code validation. This paper presents a CFD analysis of a stratified air-water flow experimental investigation performed at the Institut de Mécanique des Fluides de Toulouse in 1985, which shares some common physical features with the ECC injection in PWR cold leg. Numerical simulations have been carried out with two commercial codes (Fluent and Ansys CFX, and a research code (NEPTUNE CFD. The aim of this work, carried out at the University of Pisa within the NURESIM IP, is to validate the free surface flow model implemented in the codes against experimental data, and to perform code-to-code benchmarking. Obtained results suggest the relevance of three-dimensional effects and stress the importance of a suitable interface drag modelling.

  20. Simulation of Radioactive Corrosion Product in Primary Cooling System of Japanese Sodium-Cooled Fast Breeder Reactor

    Science.gov (United States)

    Matuo, Youichirou; Miyahara, Shinya; Izumi, Yoshinobu

    Radioactive Corrosion Product (CP) is a main cause of personal radiation exposure during maintenance with no breached fuel in fast breeder reactor (FBR) plants. The most important CP is 54Mn and 60Co. In order to establish techniques of radiation dose estimation for radiation workers in radiation-controlled areas of the FBR, the PSYCHE (Program SYstem for Corrosion Hazard Evaluation) code was developed. We add the Particle Model to the conventional PSYCHE analytical model. In this paper, we performed calculation of CP transfer in JOYO using an improved calculation code in which the Particle Model was added to the PSYCHE. The C/E (calculated / experimentally observed) value for CP deposition was improved through use of this improved PSYCHE incorporating the Particle Model. Moreover, among the percentage of total radioactive deposition accounted for by CP in particle form, 54Mn was estimated to constitute approximately 20 % and 60Co approximately 40 % in the cold-leg region. These calculation results are consistent with the measured results for the actual cold-leg piping in the JOYO.

  1. Experiment data report for semiscale Mod-1 Test S-06-5. (LOFT counterpart test). [PWR

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-06-01

    Recorded test data are presented for Test S-06-5 of the Semiscale Mod-1 LOFT counterpart test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-5 was conducted from initial conditions of 2272 psia and 536/sup 0/F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold legs of the intact and broken loops to simulate emergency core coolant injection in a PWR. The purpose of Test S-06-5 was to assess the influence of the break nozzle geometry on core thermal and system response and on the subcooled and low quality mass flow rates at the break locations.

  2. System-Level Heat Transfer Analysis, Thermal- Mechanical Cyclic Stress Analysis, and Environmental Fatigue Modeling of a Two-Loop Pressurized Water Reactor. A Preliminary Study

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish [Argonne National Lab. (ANL), Argonne, IL (United States); Soppet, William [Argonne National Lab. (ANL), Argonne, IL (United States); Majumdar, Saurin [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, Ken [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-01-03

    This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in April 2015 under the work package for environmentally assisted fatigue under DOE's Light Water Reactor Sustainability program. In this report, updates are discussed related to a system level preliminary finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent thermal-mechanical stress analysis were performed for typical design-basis thermal-mechanical fatigue cycles. The in-air fatigue lives of components, such as the hot and cold legs, were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in NUREG-6909. The discussed models and results are very preliminary. Further advancement of the discussed model is required for more accurate life prediction of reactor components. This report only presents the work related to finite element modelling activities. However, in between multiple tensile and fatigue tests were conducted. The related experimental results will be presented in the year-end report.

  3. Experiment data report for LOFT nonnuclear Test L1-4. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Batt, D. L.

    1977-07-01

    Test L1-4 was the fourth in a series of five nonnuclear isothermal blowdown tests conducted by the Loss of Fluid Test (LOFT) Program. Test L1-4 was the first Nuclear Regulatory Commission standard problem (International Problem No. 5 and U.S. Problem No. 7) experiment conducted at LOFT. Data from this test will be compared with predictions generated by the standard problem participants. For this test the LOFT Facility was configured to simulate a loss-of-coolant accident in a large pressurized water reactor resulting from a 200% double-ended offset shear break in a cold leg of the primary coolant system. A hydraulic core simulator assembly was installed in place of the nuclear core. The initial conditions in the primary coolant system intact loop were temperature at 279/sup 0/C, gauge pressure at 15.65 MPa, and intact loop flow at 268.4 kg/s. During system depressurization into a simulated containment, emergency core cooling water was injected into the primary coolant system cold leg to provide data on the effects of emergency core cooling on system thermalhydraulic response.

  4. Research on the improvement of nuclear safety -The development of LOCA analysis codes for nuclear power plant-

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Won Pyo; Jung, Yung Jong; Kim, Kyung Doo; Jung, Jae Joon; Kim, Won Suk; Han, Doh Heui; Hah, Kooi Suk; Jung, Bub Dong; Lee, Yung Jin; Hwang, Tae Suk; Lee, Sang Yong; Park, Chan Uk; Choi, Han Rim; Lee, Sang Jong; Choi, Jong Hoh; Ban, Chang Hwan; Bae, Kyoo Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The present research aims at development of both a best estimate methodology on LOCA analysis and, as an application, performance analyses of safety systems. SBLOCA analyses have been continued to examine the capacity reduction effect of ECCS since the second project year. As a results, core uncovery, which is requirement of URD has not been occurred in 6`` cold leg break. Although core uncovery has been predicted when DVI line has been broken for DVI+4-Train HPIS, the calculated PCT has lied well within the criterion. The effect of safety injection position and SIT characteristics are also analyzed for LBLOCA. The results show that cold leg injection is the most effective way and the adaption of advanced SIT could lead to elimination of LPSI pump from the safety system. On the other hand, the quantified uncertainties obtained from THTF and FLECHT/SEASET which represents blowdown and reflood phenomena, respectively, have been confirmed using IET(LOFT test). The application uncertainty for Kori unit 3 has been analyzed. Finally, application of the best estimate methodology using the uncertainties concerned with the code, the bais, and the application, leads to overall uncertainty of about 200K for Kori unit 3. 244 figs, 22 tabs, 92 refs. (Author).

  5. Experiment data report for Semiscale Mod-1 Tests S-28-7, S-28-9, and S-28-12. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Esparza, V.; Collins, B.L.; Sackett, K.E.; Coppin, C.E.

    1978-02-01

    Recorded test data are presented for Tests S-28-7, S-28-9, and S-28-12 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Tests S-28-7, S-28-9, and S-28-12 were conducted from initial conditions of 15 736 kPa and 557 K, 15 754 kPa and 556 K, and 15 704 kPa and 559 K, respectively, to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. The specific objective of these tests was to refine the definition of the upper limit of steam generator tube ruptures at which high peak cladding temperatures occur, as set by Test S-28-1. During these tests, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant in a PWR. Thirty (Test S-28-7), 34 (Test S-28-9), and 20 (Test S-28-12) steam generator tube ruptures were simulated by a controlled injection from a heated accmulator into the intact loop hot leg.

  6. An Overview of the Pressurized Thermal Shock Issue in the Context of the NURESIM Project

    Directory of Open Access Journals (Sweden)

    D. Lucas

    2009-01-01

    Full Text Available Within the European Integrated Project NURESIM, the simulation of PTS is investigated. Some accident scenarios for Pressurized Water Reactors may cause Emergency Core Coolant injection into the cold leg leading to PTS situations. They imply the formation of temperature gradients in the thick vessel walls with consequent localized stresses and the potential for propagation of possible flaws present in the material. This paper focuses on two-phase conditions that are potentially at the origin of PTS. It summarizes recent advances in the understanding of the two-phase phenomena occurring within the geometric region of the nuclear reactor,that is, the cold leg and the downcomer, where the “PTS fluid-dynamics" is relevant. Available experimental data for validation of two-phase CFD simulation tools are reviewed and the capabilities of such tools to capture each basic phenomenon are discussed. Key conclusions show that several two-phase flow subphenomena are involved and can individually be simulated at least at a qualitative level, but the capability to simulate their interaction and the overall system performance is still limited. In the near term, one may envisage a simplified treatment of two-phase PTS transients by neglecting some effects which are not yet well controlled, leading to slightly conservative predictions.

  7. Experiment data report for Semiscale Mod-1 tests S-05-2A and S-05-2B (alternate ECC injection tests)

    Energy Technology Data Exchange (ETDEWEB)

    Patton, Jr., M. L.; Collins, B. L.; Sackett, K. E.

    1977-04-01

    Recorded test data are presented for Tests S-05-2A and S-05-2B of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Tests S-05-2A and S-05-2B were conducted from initial conditions of 2263 psia and 543/sup 0/F and 2272 psia and 542/sup 0/F, respectively, to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the tests, cooling water was injected into the intact loop pump suction and broken loop cold leg to simulate emergency core coolant injection in a PWR with flow rates based on system volume scaling. For Test S-05-2A the intact loop pump speed was held constant throughout the test at the initial blowdown value. During Test S-05-2B the pump speed was reduced and stopped according to a predetermined coastdown schedule.

  8. CATHARE Assessment of PACTEL LOCA Experiments with Accident Management

    Directory of Open Access Journals (Sweden)

    Luben Sabotinov

    2010-01-01

    Full Text Available This paper summarizes the analysis results of three PACTEL experiments, carried out with the advanced thermal-hydraulic system computer CATHARE 2 code as a part of the second work package WP2 (analytical work of the EC project “Improved Accident Management of VVER nuclear power plants” (IMPAM-VVER. The three LOCA experiments, conducted on the Finnish test facility PACTEL (VVER-440 model, represent 7.4% cold leg breaks with combination of secondary bleed and primary bleed and feed and different actuation modes of the passive safety injection. The code was used for both defining and analyzing the experiments, and to assess its capabilities in predicting the associated complex VVER-related phenomena. The code results are in reasonable agreement with the measurements, and the important physical phenomena are well predicted, although still further improvement and validation might be necessary.

  9. Two-Phase Flow Simulations for PTS Investigation by Means of Neptune_CFD Code

    Directory of Open Access Journals (Sweden)

    Fabio Moretti

    2008-11-01

    Full Text Available Two-dimensional axisymmetric simulations of pressurized thermal shock (PTS phenomena through Neptune_CFD module are presented aiming at two-phase models validation against experimental data. Because of PTS complexity, only some thermal-hydraulic aspects were considered. Two different flow configurations were studied, occurring when emergency core cooling (ECC water is injected in an uncovered cold leg of a pressurized water reactor (PWR—a plunging water jet entering a free surface, and a stratified steam-water flow. Some standard and new implemented models were tested: modified turbulent k-ε models with turbulence production induced by interfacial friction, models for the drag coefficient, and interfacial heat transfer models. Quite good agreement with experimental data was achieved with best performing models for both test cases, even if a further improvement in phase change modelling would be suitable for nuclear technology applications.

  10. Simple mixing model for pressurized thermal shock applications

    Energy Technology Data Exchange (ETDEWEB)

    Chexal, B.; Chao, J.; Nickell, R.; Griesbach, T. (Electric Power Research Inst., Palo Alto, CA (USA))

    1983-02-01

    The phenomenon of fluid/thermal mixing in the cold leg and downcomer of a Pressurized Water Reactor (PWR) has been a critical issue related to the concern of pressurized thermal shock. The question of imperfect mixing arises when the possibility of cold emergency core cooling water contacting the vessel wall during an overcooling transient could produce thermal stresses large enough to initiate a flaw in a radiation embrittled vessel wall. The temperature of the fluid in contact with the vessel wall is crucial to a determination of vessel integrity since temperature affects both the stresses and the material toughness of the vessel material. A simple mixing model is described which was developed as part of the EPRI pressurized thermal shock program for evaluation of reactor vessel integrity.

  11. A neural networks based ``trip`` analysis system for PWR-type reactors; Um sistema de analise de ``trip`` em reatores PWR usando redes neuronais

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Antonio Carlos Pinto Dias

    1993-12-31

    The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients`inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author) 11 refs., 39 figs.

  12. Scaling of the direct ECC bypass during LBLOCA reflood phase with direct vessel injection system

    Energy Technology Data Exchange (ETDEWEB)

    Yun, B.J.; Kwon, T.S.; Song, C.H.; Jeong, J.J. [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of); Cho, H.K.; Park, G.C. [Seoul National Univ., Dept. of Nuclear Engineering (Korea, Republic of)

    2001-07-01

    As one of the advanced design features of the Korea next generation reactor, direct vessel injection (DVI) system is being considered instead of conventional cold leg injection (CLI) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood period of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is under progress. In this paper, a new scaling method, using time and velocity reduced linear scaling law, is suggested for the design of a scaled-down experimental facility to investigate the direct ECC bypass phenomena in PWR downcomer. (authors)

  13. TRAC-PF1/MOD1 post-test calculations of the OECD LOFT Experiment LP-SB-3

    Energy Technology Data Exchange (ETDEWEB)

    Allen, E J; Neill, A P [UKAEA Atomic Energy Establishment, Winfrith (UK)

    1990-04-01

    Analysis of the small, cold leg break, OECD LOFT Experiment LP-SB-3 using the best-estimate computer code TRAC-PF1/MOD1 is presented. Descriptions of the LOFT facility and the LP-SB-3 experiment are given and development of the TRAC-PF1/MOD1 input model is detailed. The calculations performed in achieving the steady state conditions, from which the experiment was initiated, and the specification of experimental boundary conditions are outlined. Results of the TRAC-PF1/MOD1 calculation are found to be generally consistent with those reported, by members of the OECD LOFT Program Review Group, in the LP-SB-3 Comparison Report.'' Overall trends with respect to pressure histories, minimum primary system mass inventory and accumulator behaviour are reasonably well reproduced by TRAC-PF1/MOD1. 17 refs., 26 figs., 3 tabs.

  14. Research and Evaluation for Passive Safety System in Low Pressure Reactor

    Directory of Open Access Journals (Sweden)

    Peng Chuanxin

    2015-01-01

    Full Text Available Low pressure reactor is a small size advanced reactor with power of 180 MWt, which is under development at Nuclear Power Institute of China. In order to assess the ability and feasibility of passive safety system, several tests have been implemented on the passive safety system (PSS test facility. During the LOCA and SBO accident, the adequate core cooling is provided by the performance of passive safety system. In addition the best-estimate thermal hydraulic code, CATHARE V2.1, has been assessed against cold leg LOCA test. The calculation results show that CATHARE is in a satisfactory agreement with the test for the steady state and transient test.

  15. Results of the first nuclear blowdown test on single fuel rods (LOC-11 Series in PBF)

    Energy Technology Data Exchange (ETDEWEB)

    Larson, J.R.; Evans, D.R.; McCardell, R.K.

    1978-01-01

    This paper presents results of the first nuclear blowdown tests (LOC-11A, LOC-11B, LOC-11C) ever conducted. The Loss-of-Coolant Accident (LOCA) Test Series is being conducted in the Power Burst Facility (PBF) reactor at the Idaho National Engineering Laboratory, near Idaho Falls, Idaho, for the Nuclear Regulatory Commission. The objective of the LOC-11 tests was to obtain data on the behavior of pressurized and unpressurized rods when exposed to a blowdown similar to that expected in a pressurized water reactor (PWR) during a hypothesized double-ended cold-leg break. The data are being used for the development and verification of analytical models that are used to predict coolant and fuel rod pressure during a LOCA in a PWR.

  16. The evaluation of steam-water heat transfer in vertical channel using Trac code

    Energy Technology Data Exchange (ETDEWEB)

    Sang Won, Lee; Han Gon, Kim; Byong Sup, Kim; Seung, Jong [Korea Electric Power Co., Taejon (Korea, Republic of)

    2001-07-01

    The safety injection system(SIS) of the Korean Next Generation Reactor (KNGR) injects water into the reactor vessel directly for the effective usage of ECC water. The injection location is 83 inches higher than cold leg centerline. Due to this geometrical characteristics, during late reflood phase in large break LOCA, complicate thermal-hydraulic phenomena different from existing cold-leg injection plant could be occurred. Among these phenomena, the steam-water heat transfer is evaluated in this paper. We have selected TRAC code for the evaluation tool because it can handle the reactor vessel in three-dimensional coordinates and is has been validated using UPTF experiments. For this evaluation, we performed steam-water interaction experiments in vertical rectangular channel. The experiments have been performed for co-current and counter-current flow, various steam velocity, various water flow rate. Water is injected from the top of the channel as a thin film. Steam is injected from the top or the bottom of channel. All the experiments are performed in the condition of atmospheric pressure and that void fraction is ranged from 0.9 to 0.95. Therefore, it can be treated as annular-mist flow regime. TRAC model has been developed for the simulation of these experiments. As a result of simulation, it can be concluded that TRAC code predicts heat transfer coefficients as much as 10 times compared to experimental results. In order to correct these differences, we modified the heat transfer correlation for annular-mist flow in TRAC code. This modified correlation can be used heat transfer in the downcomer only. We will perform the larger break LOCA sensitivity analyses for the effect of heat transfer in the downcomer. (authors)

  17. Experiment data report for Semiscale Mod-1 Test S-02-8; blowdown heat transfer test. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Crapo, H.S.; Sackett, K.E.

    1976-08-01

    Recorded test data are presented for Test S-02-8 of the Semiscale Mod-1 blowdown heat transfer test series. This test is one of several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor system, and to provide a data base for a regulatory standard problem. Test S-02-8 was conducted from an initial cold leg fluid temperature of 542/sup 0/F and an initial pressure of 2,262 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization transient with full core power (1.6 MW). An electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core. System flow was set to achieve the full design core temperature differential of 66/sup 0/F. The flow resistance of the intact loop was based on core area scaling. During system depressurization, core power was reduced from the initial level of 1.6 MW to simulate the surface heat flux response of nuclear fuel rods until such time that departure from nucleate boiling occurs. Blowdown to the pressure suppression system was accomplished without simulated emergency core cooling injection or pressure suppression system coolant spray. The purpose of the report is to make available the uninterpreted data from Test S-02-8 for future data analysis andtest results reporting activities. The data, presented in the form of graphs in engineering units, have been analyzed only to the extent necessary to assure that they are reasonable and consistent.

  18. Analysis of Post-LOCA Core Inlet Blockage to Evaluate In-vessel Downstream Effect in APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Seok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-10-15

    The method was developed to have a conservatism to cover the uncertainty of analysis and the acceptance is judged by the representative bounding estimation. However, the important safety parameters such as the available driving head need to be confirmed by the plant specific calculation. Also an interaction between the debris induced head loss and the core flow rate needs to be explained because the head loss induced by debris in actual condition may reduce the core inflow rate faster. To confirm the safety parameters, in this study, thermal-hydraulic response considering the core inlet blockage (CIB) by debris during LTCC process following a double-ended guillotine break of cold leg (CLB), one of hot leg (HLB) and one of intermediate leg (ILB) of the APR1400 were calculated, respectively. MARS-KS 1.3 code has been used. The CIB has been modeled by the closure of valves to the core in exponential manner with time to observe the behavior near the complete blockage. To understand the effect of core inlet blockage (CIB) during a long term core cooling (LTCC) phase following a loss-of-coolant accident (LOCA) in the light of in-vessel downstream effect (IDE) of Generic Safety Issue (GSI) 191, double-ended guillotine break of hot leg (HLB), one of cold leg (CLB) and one of intermediate leg (ILB) were calculated, respectively. And the important safety parameters such as the available driving head and the head loss due to debris were calculated using MARS-KS code and discussed in comparison with the WCAP method. As a result, a little delayed heatup behavior of the fuel cladding was found for all the cases, which due to the redistribution of flow within the core after blockage.

  19. Scaling analysis of the thermal-hydraulic test facility for the large break LOCA of KNGR

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Byong Jo; Kwon, Tae Soon; Song, Chul Hwa; Euh, Dong Jin; Chu, In Cheol; Cho, Hyoung Kyu; Park, Jong Kyun

    2001-03-01

    Korea Next Generation Reactor(KNGR) adopts a Direct Vessel Injection (DVI) system instead of conventional Cold Leg Injection (CLI) system. In this report, a scaling analysis for the steam-water test facility of KNGR with DVI under reflood phase of Loss of Coolant Accident(LBLOCA) is carried out. The major objectives of the test facility are to clarify the thermal hydraulics phenomena in the upper downcomer region and to provide experimental data for evaluating or validating relevant thermal hydraulic models and correlations of the best estimate codes. The test facility should be designed based on the appropriate scaling law so that the same thermal hydraulics phenomena is happened as in the case of prototype. For these, the investigations of previous scaling laws are carried out. And, in the present study, a new scaling approach, named the modified linear scaling, is developed for the design of a scaled-down experimental facility. Its velocity is scaled by a Wallis-type parameter and an aspect ratio of experimental facility is preserved with that of a prototype. The test facility is designed primarily by a volume scaling law and the area ratio of test facility is set to be 1/24.3. However, additional DVI nozzles are also installed at the elevation which is determined by the modified linear scaling law. It is for the scaling analysis of ECC bypass fraction. The cold leg, hot leg and DVI nozzles are additionally attached in the upper annulus downcomer region so that the UPTF counterpart test is possible.

  20. Experimental Study of the APR+ Direct ECC Bypass in the Air-water Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kihwan; Choi, Hae-Seob; Park, Kil-won; Kwon, Tae-Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The APR+ is an improved Korean Nuclear Power Reactor, which has been developed as a two loop evolutionary PWR (Pressure Water Reactor) with a number of advanced design features to enhance safety based on the APR-1400 technology. The emergency core cooling system (ECC) of the APR+ is different with that of the APR-1400, though the APR+ adopted a direct vessel injection (DVI) system which is the same design features of the APR-14000. The main difference of the DVI+ is the emergency core barrel duct (ECBD) which is designed to increase the amount of the injection water to the core region. The performance of the DVI system has been an important issues for past decades, and many researchers have studied the related thermal-hydraulic technical issues such as the ECC bypass fraction, the steam condensation effect, temperature distribution, sub-cooling margin, and etc. However, the previous research cannot be directly applicable to the APR+ owing to the unique features of the DVI+. The current study will elaborate on the experimental evaluation of the direct ECC bypass performance. The 1/5 ECC bypass test facility which is designed with a linearly reduced 1/5 scale referring to the APR+ was used to investigate the effect of the DVI+ injection nozzle location and the broken cold leg velocity on the direct ECC bypass fraction. However, air is used as a working fluid to simulate the steam flow induced from the broken cold leg, and thus, the direct contact condensation effect is not considered in this study. Experimental study for the direct ECC bypass phenomena has been carryout out with various the injection mode and air velocity conditions. The tests were performed in the 1/5 scale ECC bypass test facility, and the test condition was defined using a scaling law referring to the APR+ reactor. Test results showed that the direct ECC bypass fraction was greatly enhanced compared with the reference test (w/o ECBD)

  1. Measurement and evaluation of Corrosion Products deposition distribution in the Experimental Fast Reactor JOYO

    Energy Technology Data Exchange (ETDEWEB)

    Aoyama, Takafumi; Sumino, Kozo [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center; Masui, Tomohiko; Saikawa, Takuya

    1997-12-01

    The Corrosion Product (CP) is the major radiation source in the primary cooling system of an LMFBR plant. It is important to characterize and predict the CP behavior to reduce the personnel exposure dose due to CP deposition. The CP measurement was carried out in the Experimental Fast Reactor JOYO during the 11th annual inspection period when the accumulated reactor thermal power reached about 143 GWd. The CP deposition density was measured using a pure germanium detector. The plastic scintillation fiber (PSF) was applied for the gamma-ray dose rate distribution measurement and compared with the thermoluminescence dosimeter (TLD). The major results obtained by the CP measurements in JOYO are the follows: (1) The major CP nuclides deposited in the primary cooling system are {sup 54}Mn and {sup 60}Co. {sup 54}Mn is the dominant isotope and it tends to deposit in the cold leg region. On the other hand, {sup 60}Co deposits mainly in the hot leg region. The deposition density of {sup 54}Mn is about seven times as much as that of {sup 60}Co in the cold leg region and twice in the hot leg region. (2) The deposition densities of {sup 54}Mn and {sup 60}Co, and the gamma-dose rate were decreased from the last data in the previous annual inspection period mainly due to the short operation time and the longer cooling time. (3) The continuous gamma-ray dose rate distribution up to 10m can be measured by using the PSF in a few minutes. The PSF is suitable to measure the gamma-ray dose rate distribution in the maintenance work area where it is narrow and the mixture of gamma-ray sources from primary pipings and components. The data base of detailed gamma-ray dose rate distribution was greatly extended by the PSF. (author)

  2. Analysis on Non-Uniform Flow in Steam Generator During Steady State Natural Circulation Cooling

    Directory of Open Access Journals (Sweden)

    Susyadi

    2007-07-01

    Full Text Available Investigation on non uniform flow behavior among U-tube in steam generator during natural circulation cooling has been conducted using RELAP5. The investigation is performed by modeling the steam generator into multi channel models, i.e. 9-tubes model. Two situations are implemented, high pressure and low pressure cases. Using partial model, the calculation simulates situation similar to the natural circulation test performed in LSTF. The imposed boundary conditions are flow rate, quality, pressure of the primary side, feed water temperature, steam generator liquid level, and pressure in the secondary side. Calculation result shows that simulation using model with nine tubes is capable to capture important non-uniform phenomena such as reverse flow, fill-and-dump, and stagnant vertical stratification. As a result of appropriate simulation of non uniform flow, the calculated steam generator outlet flow in the primary loop is stable as observed in the experiments. The results also clearly indicate the importance of simulation of non-uniform flow in predicting both the flow stability and heat transfer between the primary and secondary side. In addition, the history of transient plays important role on the selection of the flow distribution among tubes. © 2007 Atom Indonesia. All rights reserved

  3. Simulation of a main steam line break with steam generator tube rupture using trace

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo, S.; Querol, A.; Verdu, G. [Departamento de Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Camino de Vera s/n, 46022, Valencia (Spain)

    2012-07-01

    A simulation of the OECD/NEA ROSA-2 Project Test 5 was made with the thermal-hydraulic code TRACE5. Test 5 performed in the Large Scale Test Facility (LSTF) reproduced a Main Steam Line Break (MSLB) with a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR). The result of these simultaneous breaks is a depressurization in the secondary and primary system in loop B because both systems are connected through the SGTR. Good approximation was obtained between TRACE5 results and experimental data. TRACE5 reproduces qualitatively the phenomena that occur in this transient: primary pressure falls after the break, stagnation of the pressure after the opening of the relief valve of the intact steam generator, the pressure falls after the two openings of the PORV and the recovery of the liquid level in the pressurizer after each closure of the PORV. Furthermore, a sensitivity analysis has been performed to know the effect of varying the High Pressure Injection (HPI) flow rate in both loops on the system pressures evolution. (authors)

  4. A hydrodynamic model of nearshore waves and wave-induced currents

    Directory of Open Access Journals (Sweden)

    Ahmed Khaled Seif

    2011-09-01

    Full Text Available In This study develops a quasi-three dimensional numerical model of wave driven coastal currents with accounting the effects of the wave-current interaction and the surface rollers. In the wave model, the current effects on wave breaking and energy dissipation are taken into account as well as the wave diffraction effect. The surface roller associated with wave breaking was modeled based on a modification of the equations by Dally and Brown (1995 and Larson and Kraus (2002. Furthermore, the quasi-three dimensional model, which based on Navier-Stokes equations, was modified in association with the surface roller effect, and solved using frictional step method. The model was validated by data sets obtained during experiments on the Large Scale Sediment Transport Facility (LSTF basin and the Hazaki Oceanographical Research Station (HORS. Then, a model test against detached breakwater was carried out to investigate the performance of the model around coastal structures. Finally, the model was applied to Akasaki port to verify the hydrodynamics around coastal structures. Good agreements between computations and measurements were obtained with regard to the cross-shore variation in waves and currents in nearshore and surf zone.

  5. The Use of System Codes in Scaling Studies: Relevant Techniques for Qualifying NPP Nodalizations for Particular Scenarios

    Directory of Open Access Journals (Sweden)

    V. Martinez-Quiroga

    2014-01-01

    Full Text Available System codes along with necessary nodalizations are valuable tools for thermal hydraulic safety analysis. Qualifying both codes and nodalizations is an essential step prior to their use in any significant study involving code calculations. Since most existing experimental data come from tests performed on the small scale, any qualification process must therefore address scale considerations. This paper describes the methodology developed at the Technical University of Catalonia in order to contribute to the qualification of Nuclear Power Plant nodalizations by means of scale disquisitions. The techniques that are presented include the so-called Kv-scaled calculation approach as well as the use of “hybrid nodalizations” and “scaled-up nodalizations.” These methods have revealed themselves to be very helpful in producing the required qualification and in promoting further improvements in nodalization. The paper explains both the concepts and the general guidelines of the method, while an accompanying paper will complete the presentation of the methodology as well as showing the results of the analysis of scaling discrepancies that appeared during the posttest simulations of PKL-LSTF counterpart tests performed on the PKL-III and ROSA-2 OECD/NEA Projects. Both articles together produce the complete description of the methodology that has been developed in the framework of the use of NPP nodalizations in the support to plant operation and control.

  6. Natural circulation in a VVER reactor geometry: Experiments with the PACTEL facility and Cathare simulations

    Energy Technology Data Exchange (ETDEWEB)

    Raussi, P.; Kainulainen, S. [Lappeenranta Univ. of Technology, Lappeenranta (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

    1995-09-01

    There are some 40 reactors based on the VVER design in use. Database available for computer code assessment for VVER reactors is rather limited. Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after VVER pressurized water reactors. Flow behaviour over a range of coolant inventories was studied with a small-break experiment. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs. The cause was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. At low primary inventories, core cooling was achieved through the boiler-condenser mode. The experiment was simulated using French thermalhydraulic system code CATHARE.

  7. Proceedings of the international meeting on thermal nuclear reactor safety. Vol. 1

    Energy Technology Data Exchange (ETDEWEB)

    None

    1983-02-01

    Separate abstracts are included for each of the papers presented concerning current issues in nuclear power plant safety; national programs in nuclear power plant safety; radiological source terms; probabilistic risk assessment methods and techniques; non LOCA and small-break-LOCA transients; safety goals; pressurized thermal shocks; applications of reliability and risk methods to probabilistic risk assessment; human factors and man-machine interface; and data bases and special applications.

  8. Combination treatment of low fluence photodynamic therapy and intravitreal ranibizumab for choroidal neovascular membrane secondary to angioid streaks in Paget′s disease - 12 month results

    Directory of Open Access Journals (Sweden)

    Varsha V Prabhu

    2011-01-01

    Full Text Available Angioid streaks also called Knapp striae are small breaks in the Bruch′s membrane and have been reported with a host of systemic diseases. Rupture of streaks or development of secondary choroidal neovascular membrane (CNVM carries a dismal visual prognosis. We report the successful treatment of CNVM secondary to Paget′s disease using low fluence photodynamic therapy (PDT and intravitreal ranibizumab.

  9. Major Achievements and Prospect of the ATLAS Integral Effect Tests

    OpenAIRE

    2012-01-01

    A large-scale thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been operated by KAERI. The reference plant of ATLAS is the APR1400 (Advanced Power Reactor, 1400 MWe). Since 2007, an extensive series of experimental works were successfully carried out, including large break loss of coolant accident tests, small break loss of coolant accident tests at various break locations, steam generator tube rupture tests, feed line ...

  10. Reduced-scale water test of natural circulation for decay heat removal in loop-type sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, T., E-mail: murakami@criepi.denken.or.jp [Central Research Institute of Electric Power Industry, 1646 Abiko, Chiba (Japan); Eguchi, Y., E-mail: eguchi@criepi.denken.or.jp [Central Research Institute of Electric Power Industry, 1646 Abiko, Chiba (Japan); Oyama, K., E-mail: kazuhiro_oyama@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 2-34-17 Jinguumae, Shibuya, Tokyo (Japan); Watanabe, O., E-mail: osamu4_watanabe@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 2-34-17 Jinguumae, Shibuya, Tokyo (Japan)

    2015-07-15

    Highlights: • The natural circulation characteristics of a loop-type SFR are examined by a water test. • The performance of decay heat removal system is evaluated using a similarity law. • The effects of flow deviation in the parallel piping of a primary loop are clarified. • The reproducibility of the natural circulation test is confirmed. - Abstract: Water tests of a loop-type sodium-cooled fast reactor have been conducted to physically evaluate the natural circulation characteristics. The water test apparatus was manufactured as a 1/10-scale mock-up of the Japan Sodium-Cooled Fast Reactor, which adopts a decay heat removal system (DHRS) utilizing natural circulation. Tests simulating a variety of events and operation conditions clarified the thermal hydraulic characteristics and core-cooling performance of the natural circulation in the primary loop. Operation conditions such as the duration of the pump flow coast-down and the activation time of the DHRS affect the natural circulation characteristics. A long pump flow coast-down cools the upper plenum of the reactor vessel (RV). This causes the loss of the buoyant force in the RV. The test result indicates that a long pump flow coast-down tends to result in a rapid increase in the core temperature because of the loss of the buoyant force. The delayed activation of the DHRS causes a decrease in the natural circulation flow rate and a temperature rise in the RV. Flow rate deviation and a reverse flow appear in the parallel cold-leg piping in some events, which cause thermal stratification in the cold-leg piping. The DHRS prevents the core temperature from fatally rise even for the most severe design-basis event, in which sodium leakage in a secondary loop of the DHRS and the opening failure of a single damper of the air cooler occur simultaneously. In the water test for the case of siphon break in the primary loop, which is one of the design extension conditions, a circulation flow consisting of ascendant

  11. [Influence of physical workload patterns and breaks on heart rate recovery].

    Science.gov (United States)

    Kadoya, Manabu; Izumi, Hiroyuki; Kubota, Makoto; Yamashita, Tsuyoshi; Kumashiro, Masaharu

    2010-01-01

    It is necessary to try to achieve quick recovery from work strain by setting adequate breaks and shortening continuous working hours to prevent the accumulation of fatigue. However, there has been no research investigating the influence of the timing and lengths of breaks on individual aerobic capacities in recovery from work strain. In this study, we set three load patterns based on the length and timing of breaks: "no breaks", "one break" and "regular small breaks". We examined the differences of the heart rate variation in the recovery time after working considering the individual aerobic capacities (VO(2)max) of ten male subjects (mean age 22.3 +/- 1.7 yr) in the case of 50 W or 100 W workloads on a bicycle ergometer. When individual aerobic capacity was not considered, the "regular small breaks" condition led to the quickest recovery to the level of the resting heart rate at 50 W workload. Not all conditions showed heart rate recovery within 30 min at 100 W workload. On the other hand, when individual aerobic capacity was considered, the "regular small breaks" condition showed the quickest recovery to the level of the resting heart rate at 50 W workload in the low aerobic capacity group (VO(2)max mean 42.2 +/- 3.7 ml/kg/min). However, in the high aerobic capacity group (VO(2)max mean 54.5 +/- 4.1 ml/kg/min), the "regular small breaks" condition resulted in the quickest recovery of the level to the resting heart rate at 100W workload. Two-way repeated measures ANOVA was performed for the recovery time with respect to the rate of increase from resting heart rate to examine the influence on heart rate recovery of physical activity loads, workload patterns and individual fitness. Physical activity loads were strongly related to the increase from resting heart rate in recovery time, and workload patterns showed that the regular small breaks condition was related to the heart rate recovery in the high fitness subjects in the case of the exercise intensity of 100 W

  12. Primary coolant sampling for activated corrosion product studies at Hanford N Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bechtold, D.B.

    1985-01-31

    A special system for sampling primary coolant at N Reactor during operation has been constructed and operated from 1977 to 1983. The basic criteria and design for solving the difficult problem of getting representative samples have been presented; this report details how the instrumentation was configured and sampling was done. Equipment and procedures were put together to allow one person to enter a radiation zone, check on 5 monitoring instruments, operate two batch instruments, gather five partitioned samples, record 26 pieces of information, annotate a strip chart and leave the zone in 30 minutes while expending 10 mRem of exposure. Additionally, the reduction of the samples' analysis, digitization of strip chart information and storage of all data on data management systems is maintained. As built, the system provides 0.3 to 1.0 gpm streams of coolant from upstream and downstream of a steam generator. The streams are cooled to 50 to 60/sup 0/C. The radiation environment averages 20 to 50 mR/hr to the worker. Instruments and special equipment for data gathering at the sampler include pH, conductance, dissolved oxygen, dissolved hydrogen and nitrogen, hot leg and cold leg coolant temperatures, particle sizing, turbidimetry, filtration, and continuous strip chart recording.

  13. Initial Characterization of V-4Cr-4Ti and MHD Coatings Exposed to Flowing Li

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A [ORNL; Pawel, Steven J [ORNL; Howell, Michael [ORNL; Moser, Jeremy L [ORNL; Garner, George Walter [ORNL; Santella, Michael L [ORNL; Tortorelli, Peter F [ORNL; Wiffen, Frederick W [ORNL; Distefano, James R [ORNL

    2009-01-01

    Conduct an experiment with flowing Li in a thermal gradient to determine the compatibility of V-4Cr-4Ti and a multi-layer electrically-insulating coating needed to reduce the magneto hydrodynamic (MHD) force in the first wall of a lithium cooled blanket. A mono-metallic V-4Cr-4Ti thermal convection loop was operated in vacuum ({approx}10{sup -5}Pa) at a maximum Li temperature of 700 C for 2,355h and Li flow rate of 2-3 cm/s. Two-layer, physical vapor deposited Y{sub 2}O{sub 3}-vanadium, electrically-insulating coatings on V-4Cr-4Ti substrates as well as uncoated tensile and sheet specimens were located in the flow path in the hot and cold legs. After exposure, specimens at the top of the hot leg showed a maximum mass loss equivalent to {approx}1.3 {micro}m of uniform metal loss. Elsewhere, small mass gains were observed on the majority of specimens that also showed an increase in hardness and room temperature yield stress and a decrease in ductility consistent with interstitial uptake. Specimens that lost mass showed a decrease in yield stress and hardness. Profilometry showed no significant thickness loss from the coatings.

  14. Initial characterization of V-4Cr-4Ti and MHD coatings exposed to flowing Li

    Energy Technology Data Exchange (ETDEWEB)

    Pint, B.A. [Materials Science and Technology Division, Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6156 (United States)], E-mail: pintba@ornl.gov; Pawel, S.J.; Howell, M.; Moser, J.L.; Garner, G.W.; Santella, M.L.; Tortorelli, P.F.; Wiffen, F.W.; DiStefano, J.R. [Materials Science and Technology Division, Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6156 (United States)

    2009-04-30

    A mono-metallic V-4Cr-4Ti thermal convection loop was operated in vacuum ({approx}10{sup -5} Pa) at a maximum Li temperature of 700 deg. C for 2355 h and Li flow rate of 2-3 cm/s. Two-layer, physical vapor deposited Y{sub 2}O{sub 3}-vanadium, electrically insulating coatings on V-4Cr-4Ti substrates as well as tensile and sheet specimens were located in the flow path in the hot and cold legs. After exposure, specimens at the top of the hot leg showed a maximum mass loss equivalent to {approx}1.3 {mu}m of metal loss. Elsewhere, small mass gains were observed on the majority of specimens resulting in an increase in hardness and room temperature yield stress and a decrease in ductility consistent with the observed uptake of N and C from the Li. Specimens that lost mass showed a decrease in yield stress and hardness. Profilometry showed no significant thickness loss from the coatings.

  15. The analysis of SCS return momentum effects on the RCS water level during mid-loop operations

    Energy Technology Data Exchange (ETDEWEB)

    swang Seo, J.; Young Yang, J.; Tack Hwang, S. [Seoul National Univ. (Korea, Republic of)

    1995-09-01

    An accurate prediction of Reactor Coolant System (RCS) water levels is of importance in the determination of allowable operating range to ensure the safety during the mid-loop operations. However, complex hydraulic phenomena induced by Shutdown Cooling System (SCS) return momentum cause different water levels from those in the loop where the water level indicators are located. This was apparantly observed at the pre-core cold hydro test of the Younggwang Nuclear Unit 3 (YGN 3) in Korea. In this study, in order to analytically understand the effect of the SCS return momentum on the RCS water level and its general trend, a model using one-dimensional momentum equation, hydraulic jump, Bernoulli equation, flow resistance coefficient, and total water volume conservation has been developed to predict the RCS water levels at various RCS locations during the mid-loop conditions and the simulation results were compared with the test data. The analysis shows that the hydraulic jump in the operating cold legs in conjunction with the momentum loss throughout the RCS is the main cause creating the water level differences at various RCS locations. The prediction results provide good explanations for the test data and show the significant effect of the SCS return momentum on the RCS water levels.

  16. NRC confirmatory AP600 safety system phase I testing in the ROSA/AP600 test facility

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, G.S.; Kukita, Yutaka; Schultz, R.R.

    1996-03-01

    The NRC confirmatory phase I testing for the AP600 safety systems has been completed in the modified ROSA (Rig of Safety Assessment) test facility located at the Japan Atomic Energy Research Institute (JAERI) campus in Tokai, Japan. The test matrix included a variety of accident scenarios covering both design and beyond-design basis accidents. The test results indicate the AP600 safety systems as reflected in ROSA appear to perform as designed and there is no danger of core heatup for the accident scenarios investigated. In addition, no detrimental system interactions nor adverse effects of non-safety systems on the safety system functions were identified. However, three phenomena of interest have been identified for further examination to determine whether they are relevant to the AP600 plant. Those three phenomena are: (1) a potential for water hammer caused by rapid condensation which may occur following the actuation of the automatic depressurization system (ADS), (2) a large thermal gradient in the cold leg pipe where cooled water returns from the passive residual heat removal system and forms a thermally stratified layer, and (3) system-wide oscillations initiating following the ADS stage 4 actuation and persisting until the liquid in the pressurizer drains and steam generation in the core becomes insignificant.

  17. A Computing Approach with the Heat-Loss Model for the Transient Analysis of Liquid Metal Natural Circulation Loop

    Directory of Open Access Journals (Sweden)

    Daogang Lu

    2014-01-01

    Full Text Available The transient behaviors of natural circulation loop (NCL are important for the system reliability under postulated accidents. The heat loss and structure thermal inertia may influence the transient behaviors of NCL greatly, so a transient analysis model with consideration of heat loss was developed based on the MATLAB/Simulink to predict the thermal-hydraulic characteristic of liquid metal NCL. The transient processes including the start-up, the loss of pump, and the shutdown of thermal-hydraulic ADS lead bismuth loop (TALL experimental facility were simulated by using the model. A good agreement is obtained to validate the transient model. The appended structure would provide significant thermal inertia and flatten the temperature distribution in the transients. The oscillations of temperature and flow rate are also weakened. The temperature difference between hot leg and cold leg would increase with the decrease of heat loss, so the flow rate increases as well. However, a significant increase of hot section temperature may cause a failure of facility integrity due to the decrease of heat loss. Hence, the full power of the core tank may also be limited.

  18. Regular postexercise cooling enhances mitochondrial biogenesis through AMPK and p38 MAPK in human skeletal muscle.

    Science.gov (United States)

    Ihsan, Mohammed; Markworth, James F; Watson, Greig; Choo, Hui Cheng; Govus, Andrew; Pham, Toan; Hickey, Anthony; Cameron-Smith, David; Abbiss, Chris R

    2015-08-01

    This study investigated the effect of regular postexercise cold water immersion (CWI) on muscle aerobic adaptations to endurance training. Eight males performed 3 sessions/wk of endurance training for 4 wk. Following each session, subjects immersed one leg in a cold water bath (10°C; COLD) for 15 min, while the contralateral leg served as a control (CON). Muscle biopsies were obtained from vastus lateralis of both CON and COLD legs prior to training and 48 h following the last training session. Samples were analyzed for signaling kinases: p38 MAPK and AMPK, peroxisome proliferator-activated receptor gamma coactivator-1α (PGC-1α), enzyme activities indicative of mitochondrial biogenesis, and protein subunits representative of respiratory chain complexes I-V. Following training, subjects' peak oxygen uptake and running velocity were improved by 5.9% and 6.2%, respectively (P 0.8) were noted with changes in protein content of p38 (d = 1.02, P = 0.064), PGC-1α (d = 0.99, P = 0.079), and peroxisome proliferator-activated receptor α (d = 0.93, P = 0.10) in COLD compared with CON. No differences between conditions were observed in the representative protein subunits of respiratory complexes II, IV, and V and in the activities of several mitochondrial enzymes (P > 0.05). These findings indicate that regular CWI enhances p38, AMPK, and possibly mitochondrial biogenesis. Copyright © 2015 the American Physiological Society.

  19. Preliminary Sensitivity Study of Upper Head Nodalization for LBLOCA in APR-1400

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Dong Gu; Yoo, Seung Hun; Cho, Dae-Hyung [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, the key-way bypass was determined to be - 0.3 %. The steady state condition which is the initial condition for LBLOCA was obtained by MARS-KS calculation. Up to now, it was assumed that the temperature of the upper dome in APR-1400 was close to that of the cold leg. However, it was found that the temperature of the upper head/dome might be a little lower than or similar to that of the hot leg through the evaluation of the detailed design data. Since the higher upper head temperature affects blowdown quenching and peak cladding temperature in the reflood phase, the nodalization for upper head should be modified. In this study, the preliminary sensitivity study of original and modified nodalization for LBLOCA was performed, and the effect of upper head nodalization and temperature was evaluated qualitatively. In this study, the preliminary sensitivity study of original and modified nodalization for upper head in APR-1400 was performed, and the effect of upper head nodalization and temperature on LBLOCA PCT was evaluated qualitatively. Through the transient calculation, it was confirmed that the upper head temperature affects the water inventory in the upper head at the early stage of LBLOCA so it does the blowdown quenching and following reflood PCT significantly. The results in this study were caused by very conservative upper head temperature determination.

  20. UPTF-TRAM experiments for SBLOCA: Evaluation of condensation processes in TRAM tests A6 and A7

    Energy Technology Data Exchange (ETDEWEB)

    Sonneburg, H.G.; Tuunanen, J.; Palazov, V.V. [Gesellschaft fuer Anlagen-und Reaktorsicherheit (GRS), Muenchen (Germany)

    1995-09-01

    The investigation of thermal-hydraulic phenomena related to reactor transients with accident management measures is the goal of the TRansient and accident Management (TRAM) experimental programme being carried out at the Upper Plenum Test Facility (UPTF) at Mannheim (Germany). These experimental investigations and test analyses are funded by the German Federal Minister for Research and Technology (BMFT). The UPTF simulates these phenomena in a 1:1 such relative to the dimension of a PWR. Condensation of steam during Emergency Core Cooling (ECC) water injection from accumulators into the primary system is one of the phenomena studied within the accumulators into the primary system is one of the phenomena studied within the TRAM programme. This phenomenon partly controls the efficiency of accumulator injection if the high pressure safety systems fail. Beside this, the condensation within the nitrogen inside the accumulator for a certain period controls the pressure development inside the accumulator. Thus, both condensation phenomena determine the ECC flow rate delivered to the primary system. Concerning the condensation inside the primary system, this is also of safety relevance in the case of Pressurized Thermal Shock (PTS) during cold leg injection.

  1. Thermal-Mechanical Stress Analysis of PWR Pressure Vessel and Nozzles under Grid Load-Following Mode: Interim Report on the Effect of Cyclic Hardening Material Properties and Pre-existing Cracks on Stress Analysis Results

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish [Argonne National Lab. (ANL), Argonne, IL (United States); Soppet, William [Argonne National Lab. (ANL), Argonne, IL (United States); Majumdar, Saurin [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, Ken [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-15

    This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable under the work package for environmentally assisted fatigue as part of DOE’s Light Water Reactor Sustainability Program. In a previous report (September 2015), we presented tensile and fatigue test data and related hardening material properties for 508 low-alloys steel base metal and other reactor metals. In this report, we present thermal-mechanical stress analysis of the reactor pressure vessel and its hot-leg and cold-leg nozzles based on estimated material properties. We also present results from thermal and thermal-mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting cracks in the reactor nozzles (axial or circumferential crack). In addition, results from validation stress analysis based on tensile and fatigue experiments are reported.

  2. Investigation on two-phase critical flow for loss-of-coolant accident of pressurized water reactor

    Institute of Scientific and Technical Information of China (English)

    1996-01-01

    The previous investigations were mainly conducted under the condition of low pressure,however,the steam-water specific volume and the interphase evaporation rate in high pressure are much different from those in low pressure,Therefore,the new experimental and theoretical investigation are performed in Xi'an Jiaotong University.The investigation results could be directly applied to the analysis of loss-of -coolant accident for pressurized water reacor.The system transition characteristics of cold leg and hot leg break loss-of -coolant tests are described for convective circulation test loop.Two types of loss-of-coolant accident are identified for :hot leg” break,while three types for “cold leg”break and the effect parameters on the break geometries.Tests indicate that the mass flow rate with convergent-divergent nozzle reaches the maximum value among the different break sections at the same inlet fluid condition because the fluid separation does not occur.A wall surface cavity nucleation model is developed for prediction of the critical mass flow rate with water flowing in convergentdivergent nozzles.

  3. Current Sheet Structures Observed by the TESIS EUV Telescope During A Flux Rope Eruption on the Sun

    CERN Document Server

    Reva, Anton; Kuzin, Ssergey

    2016-01-01

    We use the TESIS EUV telescope to study the current sheet signatures observed during flux rope eruption. The special feature of the TESIS telescope was its ability to image the solar corona up to a distance of 2 $R_\\odot$ from the Sun's center in the Fe 171 \\AA\\ line. The Fe 171 \\AA\\ line emission illuminates the magnetic field lines, and the TESIS images reveal the coronal magnetic structure at high altitudes. The analyzed CME had a core with a spiral-flux rope-structure. The spiral shape indicates that the flux rope radius varied along its length. The flux rope had a complex temperature structure: cold legs (70 000 K, observed in He 304 \\AA\\ line) and a hotter core (0.7 MK, observed in Fe 171 \\AA\\ line). Such structure contradicts the common assumption that the CME core is a cold prominence. When the CME impulsively accelerated, a dark double Y-structure appeared below the flux rope. The Y-structure timing, location, and morphology agree with the previously performed MHD simulations of the current sheet. We...

  4. Current Sheet Structures Observed by the TESIS EUV Telescope during a Flux Rope Eruption on the Sun

    Science.gov (United States)

    Reva, A. A.; Ulyanov, A. S.; Kuzin, S. V.

    2016-11-01

    We use the TESIS EUV telescope to study the current sheet signatures observed during flux rope eruption. The special feature of the TESIS telescope was its ability to image the solar corona up to a distance of 2 {R}⊙ from the Sun’s center in the Fe 171 Å line. The Fe 171 Å line emission illuminates the magnetic field lines, and the TESIS images reveal the coronal magnetic structure at high altitudes. The analyzed coronal mass ejection (CME) had a core with a spiral—flux rope—structure. The spiral shape indicates that the flux rope radius varied along its length. The flux rope had a complex temperature structure: cold legs (70,000 K, observed in He 304 Å line) and a hotter core (0.7 MK, observed in Fe 171 Å line). Such a structure contradicts the common assumption that the CME core is a cold prominence. When the CME impulsively accelerated, a dark double Y-structure appeared below the flux rope. The Y-structure timing, location, and morphology agree with the previously performed MHD simulations of the current sheet. We interpreted the Y-structure as a hot envelope of the current sheet and hot reconnection outflows. The Y-structure had a thickness of 6.0 Mm. Its length increased over time from 79 Mm to more than 411 Mm.

  5. Design optimisation of a nanofluid injection system for LOCA events in a nuclear power plant

    Science.gov (United States)

    Călimănescu, I.; Stan, L. C.; Velcea, D. D.

    2016-08-01

    The safety issues inside a Nuclear Power Plant (NPP) are encompassing their capacity to ensure the heat sink, meaning the capacity of the systems to release the heat from the rector to the environment. The nanofluids having good heat transfer properties, are recommended to be used in such applications. The paper is solving the following scenario: considering the Safety Injection tank and the Nanofluid injection Tank, and considering the Nanofluid injection Tank filled with a 10% alumina-water nanofluid, how can we select the best design of the connecting point between the pipes of the SIT and the Nanofluid Tank and the pressures inside of any of these tanks in order to have the biggest density of nanoparticles leaving the tanks toward the cold leg. In conclusion the biggest influence over the rate of disposal of the nanofluid inside ECCS is that of the pressure inside the SIT followed in order by the injection pipe diameter and the pressure inside the nanofluid tank. The optimum balance of these three design parameters may be reached following the procedure shown in this paper.

  6. RELAP5 Calculations of Bethsy 9.1b Test

    Directory of Open Access Journals (Sweden)

    Andrej Prošek

    2012-01-01

    Full Text Available Recently, several advanced computational tools for simulating reactor system behavior during real and hypothetical transient scenarios were developed. The TRAC/RELAP Advanced Computational Engine (TRACE is the latest in a series of advanced, best-estimate reactor system codes developed by the United States Nuclear Regulatory Commission (US NRC. Nevertheless, the RELAP5/MOD3.3 computer code will be maintained in the next years. The purpose of the present study was to assess how the accuracy of Bethsy 9.1b test calculation depends on the US NRC RELAP5 code version used. Bethsy 9.1b test (International Standard Problem no. 27 was 5.08 cm equivalent diameter cold leg break without high-pressure safety injection and with delayed ultimate procedure. Seven different RELAP5 code versions were used and as much as possible the same input model. The obtained results indicate that the results obtained by the oldest and latest RELAP5 versions are in general comparable for Bethsy 9.1b test. This is very important for the validity of the results, obtained in the past with older RELAP5 versions. Due to the fact that observation was restricted to Bethsy 9.1b posttest, with its own physical phenomena, this conclusion could be generalized only for scenarios having similar range of the considered Bethsy transient conditions.

  7. The sensitivity analysis for APR1400 nodalization under Large Break LOCA condition based on mars code

    Directory of Open Access Journals (Sweden)

    Jang Hyung-Wook

    2017-01-01

    Full Text Available The phenomena of loss of coolant accident have been investigated for long time and the result of experiment shows that the flow condition in the downcomer during the end-of-blowdown were highly multi-dimensional at full-scale. However, the downcomer nodalization of input deck for large break loss of coolant accident used in advanced power reactor 1400 analyses are made up with 1-D model and improperly designed to describe realistic coolant phenomena during loss of coolant accident analysis. In this paper, the authors modified the nodalization of MARS code LBLOCA input deck and performed LBLOCA analysis with new input deck. From original LBLOCA input deck file, the nodalization of downcomer and junction connections with 4 cold legs and direct vessel injection lines are modified for reflecting the realistic cross-flow effect and real downcomer structure. The analysis results show that the peak cladding temperature of new input deck decreases more rapidly than previous result and that the drop of peak cladding temperature was advanced by application of momentum flux term in cross-flow. Additionally, the authors developed a new input deck with multi-dimensional downcomer model and ran MARS code with multi-dimensional input deck as well. By using the modified input deck, the Emergency core cooling system by-pass flow phenomena is better characterized and found to be consistent with both experimental report and regulatory guide.

  8. VICTORIA-92 pretest analyses of PHEBUS-FPT0

    Energy Technology Data Exchange (ETDEWEB)

    Bixler, N.E.; Erickson, C.M.

    1994-01-01

    FPT0 is the first of six tests that are scheduled to be conducted in an experimental reactor in Cadarache, France. The test apparatus consists of an in-pile fuel bundle, an upper plenum, a hot leg, a steam generator, a cold leg, and a small containment. Thus, the test is integral in the sense that it attempts to simulate all of the processes that would be operative in a severe nuclear accident. In FPT0, the fuel will be trace irradiated; in subsequent tests high burn-up fuel will be used. This report discusses separate pretest analyses of the FPT0 fuel bundle and primary circuit have been conducted using the USNRC`s source term code, VICTORIA-92. Predictions for release of fission product, control rod, and structural elements from the test section are compared with those given by CORSOR-M. In general, the releases predicted by VICTORIA-92 occur earlier than those predicted by CORSOR-M. The other notable difference is that U release is predicted to be on a par with that of the control rod elements; CORSOR-M predicts U release to be about 2 orders of magnitude greater.

  9. RELAP-5/MOD 3.2 Assessment Using an 11% Upper Plenum Break Experiment in the PSB Facility

    Energy Technology Data Exchange (ETDEWEB)

    Paul D. Bayless

    2003-01-01

    The RELAP/MOD3.2 computer code has been assessed using an 11% upper plenum break experiment in the PSB test facility at the Electrogorsk Research and Engineering Center. This work was performed as part of the U.S. Department of Energy's International Nuclear Safety Program, and is part of the effort addressing the capability of the RELAP5/MOD3.2 code to model transients in Soviet-designed reactors. Designated VVER Standard Problem PSBV1, the test addressed several important phenomena related to VVER behavior that the code needs to simulate well. The code was judged to reasonably model the phenomena of two-phase flow natural circulation in the primary coolant system, asymmetric loop behavior, leak flow, loop seal clearance in the cold legs, heat transfer in a covered core, heat transfer in a partially covered core, pressurizer thermal-hydraulics, and integral system effects. The code was judged to be in minimal agreement with the experiment data for the mixture level and entrainment in the core, leading to a user recommendation to assess the sensitivity of transient calculations to the interphase drag modeling in the core. No judgments were made for the phenomena of phase separation without mixture level formation, mixture level and entrainment in the steam generators, pool formation in the upper plenum, or flow stratification in horizontal pipes because either the phenomenon did not occur in the test or there were insufficient measurements to characterize the behavior.

  10. Optimizing steam flood performance utilizing a new and highly accurate two phase steam measurement system

    Energy Technology Data Exchange (ETDEWEB)

    Huff, B. D.; Warren, P. B. [CalResources LLC (Canada); Whorff, F. [ITT Barton (Canada)

    1995-11-01

    The development of a two phase steam measurement system was documented. The system consists of a `V` cone differential pressure device and a vortex meter velocity device in series through which the steam flows. Temperature and pressure sensors are electronically interfaced with a data logging system. The design was described as being very simple and rugged, consequently, well suited to monitoring in the field.. Steam quality measurements were made in the Kern River Field and the Coalinga Field thermal projects using a surface steam separator. In steam flood operations, steam cost is very high, hence appropriate distribution of the steam can result in significant cost reduction. This technology allows the measurement of steam flow and quality at any point in the steam distribution system. The metering system`s orifice meter was found to have a total average error of 45%, with 25% of that attributable to `cold leg` problem. Installation of the metering system was expected to result in a steam use reduction of 8%, without any impact on production. Steam re-distribution could result in a potential oil production increase of 10%. 12 refs., 8 tabs., 9 figs.

  11. Application of UPTF data for modeling liquid draindown in the downcomer region of a PWR using RELAP5/MOD2-B&W

    Energy Technology Data Exchange (ETDEWEB)

    Wissinger, G.; Klingenfus, J. [B & W Nuclear Technologies, Lynchburg, VA (United States)

    1995-09-01

    B&W Nuclear Technologies (BWNT) currently uses an evaluation model that analyzes large break loss-of-coolant accidents in pressurized water reactors using several computer codes. These codes separately calculate the system performance during the blowdown, refill, and reflooding phases of the transient. Multiple codes are used, in part, because a single code has been unable to effectively model the transition from blowdown to reflood, particularly in the downcomer region where high steam velocities do not allow the injected emergency core cooling (ECC) liquid to penetrate and begin to refill the vessel lower plenum until after the end of blowdown. BWNT is developing a method using the RELAP5/MOD2-B&W computer code that can correctly predict the liquid draindown behavior in the downcomer during the late blowdown and refill phases. Benchmarks of this method have been performed against Upper Plenum Test Facility (UPTF) data for ECC liquid penetration and valves using both cold leg and downcomer ECC injection. The use of this new method in plant applications should result in the calculation of a shorter refill period, leading to lower peak clad temperature predictions and increased core peaking. This paper identifies changes made to the RELAP/MOD2-B&W code to improve its predictive capabilities with respect to the data obtained in the UPTF tests.

  12. Analysis of fluid mixing characteristics in reactor vessel downcomer using Theofanous and Wallis` mixing model. (DVI Case)

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Bong Hyun; Kim, Hwan Yeol; Kang, Hyung Seok; Bae, Yoon Young

    1997-05-01

    Direct injection of emergency core cooling water into the reactor vessel downcomer annulus (DVI) is an unique feature of the four-train safety injection system of Korean Next Generation Reactor(KNGR). In this study, in order to evaluate the fluid mixing characteristics of the injected water for DVI case, we have suggested for application to DVI, Theofanous` regional mixing model and Wallis` experiments of flow regimes for injection water to the annulus. Theofanous`model was developed as a fluid mixing model in reactor vessel downcomer for the case of Cold Leg Injection(CLI). We have established a procedure for calculating fluid mixing temperature, calculated the mixing temperature for SBLOCA and MSLB, and compared them to those of CLI. In general, the fluid temperatures across the reactor vessel beltline are higher than 110 deg F, the RT{sub NDT} of EOL for reactor vessel material, and the values are within the acceptable limits of PTS concern. (author). 6 tabs., 21 figs., 11 refs.

  13. Advanced light water reactor plants System 80+{trademark} design certification program. Annual progress report, October 1, 1994--September 30, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    The purpose of this report is to provide the status of the progress that was made towards Design Certification of System 80+{trademark} during the US government`s 1995 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW (1350 MWe) Pressurized Water Reactor (PWR). The design covers an essentially complete plant. It is based on EPRI ALWR Utility Requirements Document (URD) improvements to the Standardized System 80 Nuclear Steam Supply System (NSSS) in operation at Palo Verde Units 1, 2, and 3. The NSSS is a traditional two-loop arrangement with two steam generators, two hot legs and four cold legs, each with a reactor coolant pump. The System 80+ standard design houses the NSSS in a spherical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual barrier to radioactivity release. Other major features include an all-digital, human-factors-engineered control room, an alternate electrical AC power source, an In-Containment Refueling Water Storage Tank (IRWST), and plant arrangements providing complete separation of redundant trains in safety systems.

  14. Passive safety system of a super fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sutanto, E-mail: sutanto@fuji.waseda.jp [Cooperative Major in Nuclear Energy, Waseda University, Tokyo (Japan); Polytechnic Institute of Nuclear Technology—National Nuclear Energy Agency, Yogyakarta (Indonesia); Oka, Yoshiaki [The University of Tokyo, Tokyo (Japan)

    2015-08-15

    Highlights: • Passive safety system of a Super FR is proposed. • Total loss of feedwater flow and large LOCA are analyzed. • The criteria of MCST and core pressure are satisfied. - Abstract: Passive safety systems of a Super Fast Reactor are studied. The passive safety systems consist of isolation condenser (IC), automatic depressurization system (ADS), core make-up tank (CMT), gravity driven cooling system (GDCS), and passive containment cooling system (PCCS). Two accidents of total loss of feedwater flow and 100% cold-leg break large LOCA are analyzed by using the passive systems and the criteria of maximum cladding surface temperature (MCST) and maximum core pressure are satisfied. The isolation condenser can be used for mitigation of the accident of total loss of feedwater flow at both supercritical and subcritical pressures. The ADS is used for depressurization leading to a loss of coolant during line switching to operation of the isolation condenser at subcritical pressure. Use of CMT during line switching recovers the lost coolant. In case of large LOCA, GDCS can be used for core reflooding. Coolant vaporization in the core released to containment through the break is condensed by passive containment cooling system. The condensate flows to the GDCS pool by gravity force. The maximum cladding surface temperature (MCST) of the accident satisfies the criterion.

  15. Thermal–mechanical stress analysis of pressurized water reactor pressure vessel with/without a preexisting crack under grid load following conditions

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov; Soppet, William K.; Majumdar, Saurin; Natesan, Krishnamurti

    2016-12-15

    Highlights: • Use of intermittent renewable-energy source in power grid is becoming a trend. • Gird load-following can leads to variable power demand from Nuclear power plant. • Reactor components can be stressed differently under gird load-following mode. • Estimation of stress–strain state under grid load-following condition is essential. - Abstract: In this paper, we present thermal–mechanical stress analysis of a pressurized water reactor pressure vessel and its hot-leg and cold-leg nozzles. Results are presented from thermal and thermal–mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting crack in the reactor nozzle (axial crack in hot leg nozzle). From the model results it is found that the stress–strain states are significantly higher in case of presence of crack than without crack. The stress–strain state under grid load following condition are more realistic compared to the stress–strain state estimated assuming simplified transients.

  16. Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, January--March 1977

    Energy Technology Data Exchange (ETDEWEB)

    Card, D. H. [ed.

    1977-06-01

    A test series to evaluate the effectiveness of different emergency core coolant injection arrangements was completed. A LOFT test series was initiated to investigate the effect of break nozzle geometry on break flow and system response during a 200 percent double-ended cold leg break test. Two analyses of results of LOFT nonnuclear experiments were made. The intact loop flow resistance on LOFT system responses during loss-of-coolant experiments was evaluated. The second in a series of five tests to evaluate the effects of fuel density, initial gap width, and fill gas composition on gap conductance was performed with four BWR-type rods tested simultaneously. A test was conducted to provide steady state, power ramp, and film boiling operation behavior data on the fuel rods constructed from irradiated and unirradiated cladding and to investigate the influence of simulated fission products on cladding integrity and performance. Cladding ballooning phenomena were also investigated. Results of preliminary postirradiation examinations of irradiation effects, power-cooling mismatch, and gap conductance tests performed previously are reported. A review of FLECHT and Semiscale forced feed and gravity feed reflood heat transfer experiments was performed. A study to predict and compare surface temperature response of Semiscale electrically heated rods with LOFT nuclear rods during a Semiscale blowdown was completed. SPERT and NSRR data were analyzed to evaluate fuel failure modes and the effect of energy deposition for cold-startup conditions.

  17. Thermal hydraulics analysis of the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-01

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

  18. The Development of a Radiation Hardened Robot for Nuclear Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Seung Ho; Kim, Chang Hoi; Seo, Yong Chil (and others)

    2007-04-15

    We has been developed two remotely controlled robotic systems. One is a underwater vehicle for inspection of the internal structures of PWRs and retrieving foreign stubs in the reactor pressure vessels and reactor coolant pipes. The other robotic system consists of a articulated-type mobile robot capable of recovering the failure of the fuel exchange machine and a mini modular mobile robot for inspection of feeder pipes with ultrasonic array sensors in PHWRs. The underwater robot has been designed by considering radiation effect, underwater condition, and accessibility to the working area. The size of underwater robot is designed to enter the cold legs. A extendable manipulator is mounted on the mobile robot, which can restore nuclear fuel exchange machine. The mini modular mobile robot is composed of dual inch worm mechanisms, which are constructed by two gripper bodies that can fix the robot body on to the pipe and move along the longitudinal and to rotate in a circumferential direction to access all of the outer surfaces of the pipe.

  19. Fission products and structural materials release, transport and containment behaviour in Phebus FPT-0 and FPT-1

    Energy Technology Data Exchange (ETDEWEB)

    Hanniet, N.; Garnier, Y.; Jacquemain, D. [Institut de Protection et de Surete Nucleaire - IPSN, Departement de Recherches en Securite - DRS, CEA Cadarache - F 13108 Saint Paul Lez Durance (France)

    1999-07-01

    The Phebus F.P. program is a wide international effort to investigate, through a series of in-pile integral experiments, LWR severe accident phenomena, in particular bundle degradation and the subsequent release and transport of radioactive materials up to the containment. Two tests simulating a low pressure cold leg break under a steam rich environment have already been successfully performed: FPT-0 in December 1993 with trace irradiated fuel and FPT-1 in July 1996 with re-irradiated BR3 fuel. Both tests have provided experimental data of high interest, particularly in the field of fission products and structural materials release from the fuel bundle, transport in the reactor coolant system (RCS) and behaviour in the containment. The analysis of FPT-1 is currently in progress, nevertheless main observations made for FPT-10 are confirmed by first FPT-1 results, i.e.: - the main mass transport phases through the RCS correspond to bundle degradation events (fuel oxidation, material re-location, pool formation); - significant amount of gaseous iodine are injected in the model containment during bundle oxidation phases; - the aerosols are multi-component with the structural materials dominant in mass; - the retention is low in the RCS pipes but aerosol deposition on containment walls is significant; - the containment sump chemistry is marked by aerosol material dissolution and the resulting iodine trapping by silver. Those results are described in some detail in the following paper. (author)

  20. Coupled RELAP5, 3D CFD and FEM analysis of postulated cracks in RPVs subjected to PTS loading

    Energy Technology Data Exchange (ETDEWEB)

    González-Albuixech, V.F., E-mail: vicente.gonzalez@psi.ch [Paul Scherrer Institut (PSI), Nuclear Energy and Safety Department, Structural Integrity Group, Villigen CH-5232 (Switzerland); Qian, G., E-mail: guian.qian@psi.ch [Paul Scherrer Institut (PSI), Nuclear Energy and Safety Department, Structural Integrity Group, Villigen CH-5232 (Switzerland); Sharabi, M. [Paul Scherrer Institut (PSI), Nuclear Energy and Safety Department, Structural Integrity Group, Villigen CH-5232 (Switzerland); Mechanical Power Engineering Department, Faculty of Engineering, Mansoura University, 35516 Mansoura (Egypt); Niffenegger, M.; Niceno, B.; Lafferty, N. [Paul Scherrer Institut (PSI), Nuclear Energy and Safety Department, Structural Integrity Group, Villigen CH-5232 (Switzerland)

    2016-02-15

    Highlights: • RPV fracture mechanics model based on RELAP5. • RPV fracture mechanics model based on CFD. • RPV fracture mechanics analysis. - Abstract: The fracture mechanic analysis of a reactor pressure vessel subjected to pressurized thermal shock loading is one of the most important issues for the assessment of life time extension of a nuclear power plant. The most severe scenario occurs during cold water injection in the cold leg due to a Loss-Of-Coolant Accident (LOCA). In the present study a comprehensive fracture mechanics analysis is performed. Two hypothetical LOCAs are assumed for an adopted reference design of a two-loop Pressurized Water Reactor. Boundary conditions obtained from the RELAP5 code are used as input for Computational Fluid Dynamics (CFD) simulations. For the structural integrity analysis, submodeling technique and the eXtended Finite Element Method (XFEM) based on temperatures calculated by CFD are applied. The results from the 3D FEM calculations are compared to those from a simplified axisymmetric model based on axisymmetric thermal hydraulic model results. The analysis identifies the worst crack orientation and location. It also proves that a complete model is needed for a correct analysis as the simplified model is not conservative and fails to describe accurately the local plume effect.

  1. RELAP/MOD3.2 Assessment Using an 11% Upper Plenum Break Experiment in the PSB Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bayless, P.D.

    2003-01-17

    The RELAP/MOD3.2 computer code has been assessed using an 11% upper plenum break experiment in the PSB test facility at the Electrogorsk Research and Engineering Center. This work was performed as part of the U.S. Department of Energy's International Nuclear Safety Program, and is part of the effort addressing the capability of the RELAP5/MOD3.2 code to model transients in Soviet-designed reactors. Designated VVER Standard Problem PSBV1, the test addressed several important phenomena related to VVER behavior that the code needs to simulate well. The code was judged to reasonably model the phenomena of two-phase flow natural circulation in the primary coolant system, asymmetric loop behavior, leak flow, loop seal clearance in the cold legs, heat transfer in a covered core, heat transfer in a partially covered core, pressurizer thermal-hydraulics, and integral system effects. The code was judged to be in minimal agreement with the experiment data for the mixture level and entrainment in the core, leading to a user recommendation to assess the sensitivity of transient calculations to the interphase drag modeling in the core. No judgments were made for the phenomena of phase separation without mixture level formation, mixture level and entrainment in the steam generators, pool formation in the upper plenum, or flow stratification in horizontal pipes because either the phenomenon did not occur in the test or there were insufficient measurements to characterize the behavior.

  2. Effectiveness of a worksite social & physical environment intervention on need for recovery, physical activity and relaxation; results of a randomized controlled trial.

    Science.gov (United States)

    Coffeng, Jennifer K; Boot, Cécile R L; Duijts, Saskia F A; Twisk, Jos W R; van Mechelen, Willem; Hendriksen, Ingrid J M

    2014-01-01

    To investigate the effectiveness of a worksite social and physical environment intervention on need for recovery (i.e., early symptoms of work-related mental and physical fatigue), physical activity and relaxation. Also, the effectiveness of the separate interventions was investigated. In this 2 × 2 factorial design study, 412 office employees from a financial service provider participated. Participants were allocated to the combined social and physical intervention, to the social intervention only, to the physical intervention only or to the control group. The primary outcome measure was need for recovery. Secondary outcomes were work-related stress (i.e., exhaustion, detachment and relaxation), small breaks, physical activity (i.e., stair climbing, active commuting, sport activities, light/moderate/vigorous physical activity) and sedentary behavior. Outcomes were measured by questionnaires at baseline, 6 and 12 months follow-up. Multilevel analyses were performed to investigate the effects of the three interventions. In all intervention groups, a non-significant reduction was found in need for recovery. In the combined intervention (n = 92), exhaustion and vigorous physical activities decreased significantly, and small breaks at work and active commuting increased significantly compared to the control group. The social intervention (n = 118) showed a significant reduction in exhaustion, sedentary behavior at work and a significant increase in small breaks at work and leisure activities. In the physical intervention (n = 96), stair climbing at work and active commuting significantly increased, and sedentary behavior at work decreased significantly compared to the control group. None of the interventions was effective in improving the need for recovery. It is recommended to implement the social and physical intervention among a population with higher baseline values of need for recovery. Furthermore, the intervention itself could be improved by increasing the

  3. Detailed CATHENA Model of the Wolsong 1 Pressure and Inventory Control System

    Energy Technology Data Exchange (ETDEWEB)

    Cha, K.H. [Korea Electric Power Research Institute, Taejon (Korea)

    2002-07-01

    The Detailed CATHENA model of Wolsong 1 is development to be able to simulate a theramal hydraulic behavior of heat transport system(HTS) Pressure and Inventory Control System(PNIC) at any power operation condition and during transient events such as mall LOCA(small loss of coolant inventory and small breaks in the primary system piping) and non-LOCA(loss of reactivity regulation, loss of flow, loss if Class IV power, loss of PNIC). (author). 12 refs., 7 figs., 6 tabs.

  4. Direct condensation and entrainment steam experiments at the Topflow-Denise facility

    Energy Technology Data Exchange (ETDEWEB)

    Seidel, Tobias [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Inst. of Fluid Dynamics

    2015-10-15

    In a hypothetical Small Break Loss of Coolant Accident (SB-LOCA) in a Pressurized Water Reactor (PWR), the Reactor Pressure Vessel wall (RPV) may be exposed to thermal stress, since the Emergency Core Cooling System (ECCS) injects cold water. The loads on the primary loop and RPV walls are determined by mixing processes with the surrounding hot water and by the condensation of steam on the surface. For the development and validation of CFD-models, experiments have to meet a high standard of reproducibility, measurement certainty and temporal and local resolution. The pressure tank technology of the TOPFLOW facility allows conducting such experiments at reasonable effort.

  5. simulation of a SGTR severe PWR-W with MELCOR code; Simulacion de un SGTR severo en un PWR-W con el codigo Melcor

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, A. J.; Jimenez Varas, G.; Israelsson, L. C.

    2014-04-01

    Steam Generator tube Rupture (SGTR) is a small break loss of coolant accident. the issues related to this kind of transients makes them different from the classics LOCA studies. SGTR accidents in Pressurized Water Reactor are known to be one of the most demanding transients for the operating crew. It this accident is not managed in a proper way it could lead to steam generator overfill and a severe accident inside containment . To simulate this accident the MELCOR code was chosen, whose aim is the assessment of the progression of severe accidents in Light Water Reactors. (Author)

  6. Qualification of a full plant nodalization for the prediction of the core exit temperature through a scaling methodology

    Energy Technology Data Exchange (ETDEWEB)

    Freixa, J., E-mail: jordi.freixa-terradas@upc.edu; Martínez-Quiroga, V., E-mail: victor.martinez.quiroga@upc.edu; Reventós, F., E-mail: francesc.reventos@upc.edu

    2016-11-15

    Highlights: • Core exit temperature is used in PWRs as an indication of core heat up. • Qualification of full scale nuclear reactors by means of a scaling methodology. • Scaling of RELAP5 calculations to full scale power plants. - Abstract: System codes and their necessary power plant nodalizations are an essential step in thermal hydraulic safety analysis. In order to assess the safety of a particular power plant, in addition to the validation and verification of the code, the nodalization of the system needs to be qualified. Since most existing experimental data come from scaled-down facilities, any qualification process must therefore address scale considerations. The Group of Thermal Hydraulic Studies at Technical University of Catalonia has developed a scaling-up methodology (SCUP) for the qualification of full-scale nodalizations through a systematic procedure based on the extrapolation of post-test simulations of Integral Test Facility experiments. In the present work, the SCUP methodology will be employed to qualify the nodalization of the AscóNPP, a Pressurized Water Reactor (PWR), for the reproduction of an important safety phenomenon which is the effectiveness of the Core Exit Temperature (CET) as an Accident Management (AM) indicator. Given the difficulties in placing measurements in the core region, CET measurements are used as a criterion for the initiation of safety operational procedures during accidental conditions in PWR. However, the CET response has some limitation in detecting inadequate core cooling simply because the measurement is not taken in the position where the cladding exposure occurs. In order to apply the SCUP methodology, the OECD/NEA ROSA-2 Test 3, an SBLOCA in the hot leg, has been selected as a starting point. This experiment was conducted at the Large Scale Test Facility (LSTF), a facility operated by the Japanese Atomic Energy Agency (JAEA) and was focused on the assessment of the effectiveness of AM actions triggered by

  7. FLECHT SEASET program. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Hochreiter, L E

    1985-11-01

    This report presents the highlights and main findings of the USNRC, EPRI, and Westinghouse cooperative FLECHT SEASET program. The report indicates areas in which the results of the program can contribute to revising the current licensing requirements for Loss of Coolant (LOCA) safety analysis for PWRs. Also identified are several technical areas in which the new FLECHT SEASET data and analysis can lead to improved safety analysis modeling, and thereby to predicted PWR response for postulated accident scenarios. Significant progress has been made in the modeling areas of nonequilibrium dispersed two-phase flow during reflood. Improved models and understanding of this rod bundle cooling regime are summarized in this report. Another important result of the FLECHT SEASET program arises from the natural circulation test series, which investigated single-phase, two-phase, and reflux condensation cooling modes of a scaled PWR under small-break LOCA conditions. The tests and subsequent analysis constitute one of few complete sets of data for these cooling modes in which full-height, multitube steam generators with sufficient instrumentation were used to examine primary-to-secondary heat transfer in the generators. It is believed that the natural circulation test data will be extremely useful to benchmark the improved post-TMI small-break LOCA computer codes. 170 figs., 13 tabs.

  8. Implementing a Nuclear Power Plant Model for Evaluating Load-Following Capability on a Small Grid

    Science.gov (United States)

    Arda, Samet Egemen

    A pressurized water reactor (PWR) nuclear power plant (NPP) model is introduced into Positive Sequence Load Flow (PSLF) software by General Electric in order to evaluate the load-following capability of NPPs. The nuclear steam supply system (NSSS) consists of a reactor core, hot and cold legs, plenums, and a U-tube steam generator. The physical systems listed above are represented by mathematical models utilizing a state variable lumped parameter approach. A steady-state control program for the reactor, and simple turbine and governor models are also developed. Adequacy of the isolated reactor core, the isolated steam generator, and the complete PWR models are tested in Matlab/Simulink and dynamic responses are compared with the test results obtained from the H. B. Robinson NPP. Test results illustrate that the developed models represents the dynamic features of real-physical systems and are capable of predicting responses due to small perturbations of external reactivity and steam valve opening. Subsequently, the NSSS representation is incorporated into PSLF and coupled with built-in excitation system and generator models. Different simulation cases are run when sudden loss of generation occurs in a small power system which includes hydroelectric and natural gas power plants besides the developed PWR NPP. The conclusion is that the NPP can respond to a disturbance in the power system without exceeding any design and safety limits if appropriate operational conditions, such as achieving the NPP turbine control by adjusting the speed of the steam valve, are met. In other words, the NPP can participate in the control of system frequency and improve the overall power system performance.

  9. Timing analysis of PWR fuel pin failures

    Energy Technology Data Exchange (ETDEWEB)

    Jones, K.R.; Wade, N.L.; Katsma, K.R.; Siefken, L.J. (EG and G Idaho, Inc., Idaho Falls, ID (United States)); Straka, M. (Halliburton NUS, Idaho Falls, ID (United States))

    1992-09-01

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

  10. Starting up of new steam generator of N4 1450 MWE plants

    Energy Technology Data Exchange (ETDEWEB)

    Bussy, B. [EDF, Villeurbanne Cedex (France); Dague, G.; Slama, G. [Framatome, Paris-le-Defense (France)

    1998-07-01

    The first N4 plant, CHOOZ B1, was commissioned in 1997. This plant of 1450 MWe capacity is equipped with four steam generators of a new design fitted with an axial economiser. The axial economiser principle essentially consists in directing all the feedwater to the cold leg of the tube bundle and about 90% of the recirculated water to the hot leg. A double wrapper alone, the downcomer allows cold water to enter tube bundle as in a boiler-type steam generator andprevents excessive vibration due to cross flow. The main result of this design is to enhance the heat exchange efficiency between the primary and the secondary sides and to increase the steam pressure as compared to a boiler steam generator having the same heat exchange area. As it is the common practice in France for the first plant unit of a new model, one of the four steam generators has been specifically instrumented in order to assess its actual operating characteristics and verify their consistency with the predicted values resulting from the design studies and from the qualification tests. The test programme and the dedicated instrumentation allowed measurement of essential S.G. thermal hydraulics parameters (saturation pressure, flow velocities, pressure drops, circulation ratios, temperature distribution,...) and assessment of vibratory behaviour of tube bundle. The results are in agreement with the tests carried out in the course of the steam generator development and qualification programme, notably a few years ago on the large 25 MWth MEGEVE test facility. They confirm adequacy of the axial economiser design. The paper describes successively: the specificities of the new steam generator design; the instrumentation and the test programme; and the main results obtained from the numerous tests performed during plant start-up. (author)

  11. Postexercise muscle cooling enhances gene expression of PGC-1α.

    Science.gov (United States)

    Ihsan, Mohammed; Watson, Greig; Choo, Hui Cheng; Lewandowski, Paul; Papazzo, Annateresa; Cameron-Smith, David; Abbiss, Chris R

    2014-10-01

    This study aimed to investigate the influence of localized muscle cooling on postexercise vascular, metabolic, and mitochondrial-related gene expression. Nine physically active males performed 30 min of continuous running at 70% of their maximal aerobic velocity, followed by intermittent running to exhaustion at 100% maximal aerobic velocity. After exercise, subjects immersed one leg in a cold water bath (10°C, COLD) to the level of their gluteal fold for 15 min. The contralateral leg remained outside the water bath and served as control (CON). Core body temperature was monitored throughout the experiment, whereas muscle biopsies and muscle temperature (Tm) measurements were obtained from the vastus lateralis before exercise (PRE), immediately postexercise (POST-EX, Tm only), immediately after cooling, and 3 h postexercise (POST-3H). Exercise significantly increased core body temperature (PRE, 37.1°C ± 0.4°C vs POST-EX, 39.3°C ± 0.5°C, P COLD legs (PRE, 34.2°C ± 0.9°C vs POST-EX, 39.4°C ± 0.3°C), respectively (P COLD (28.9°C ± 2.3°C vs 37.0°C ± 0.8°C, P COLD at POST-3H (P = 0.014). Significant time effects were evident for changes in vascular endothelial growth factor (P = 0.038) and neuronal nitric oxide synthase (P = 0.019) expression. However, no significant condition effects between COLD and CON were evident for changes in both vascular endothelial growth factor and neuronal nitric oxide synthase expressions. These data indicate that an acute postexercise cooling intervention enhances the gene expression of PGC-1α and may therefore provide a valuable strategy to enhance exercise-induced mitochondrial biogenesis.

  12. 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.

  13. CIEMAT Contribution to the PHEBEN-2 Project: Interpretation of the PHEBUS-FPT1. Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.; Pra, C. l. del; Rincon, A. M.

    2003-07-01

    This report summarises the CIEMAT contribution to the analysis of the FPT1 test of the PHEBUS-FP Project. The work carried out has been focussed on containment phenomena. the thermal hydraulic evolution and the aerosol behaviour has been simulated with CONTAIN 2.0 code, whereas the iodine chemistry has been modelled with IODE 4.2. In both cases a number of hypotheses and approximations have been adopted. The FPT1 experiment investigated core degradation and release, transport and behaviour of fission products and aerosols under the anticipated conditions for a low pressure accident sequence with a pipe break at the cold leg. The containment scenario was essentially characterised by a condensing and unsaturated atmosphere in contact with an acidic sump. CONTAIN 2.0 has provided an accurate picture of the thermo-hydraulic and aerosol behaviour, whereas IODE 4.2, although succeeded in predicting the overall iodine mass distribution, but it has been unable to capture the gaseous iodine evolution during the experiment. Steam input and condensation determined the thermal conditions of the vessel and made around 30% of particulate mass deplete onto condensing structures. Most of iodine was trapped by silver from the control rods and formed AgI in the sump. the deviations of predictions regarding gaseous iodine point out the need of further development of organics-iodine interaction models. Finally, it should be underlined that the simulation has shed light on experimental aspects as well. The measured steam input history should have been slightly different from the one specified in the final report: a new profile has been proposed. The samplings of airborne caesium (rather different from the {gamma}-spectrometry data) are the most reliable measurements. (Author) 10 refs.

  14. 船用堆大破口失水事故放射性后果分析%Radioactive Consequence Analysis on Large-break LOCA of Marine Reactor

    Institute of Scientific and Technical Information of China (English)

    王伟; 陈力生; 张帆; 刘海鹏

    2015-01-01

    Based on the severe accident analysis program MELCOR as computational tool ,computational model of the marine reactor was established in the paper .Both the source behavior and radioactive consequence were researched when the double‐ended rupture large‐break LOCA happened on the cold leg of a typical ship reactor . T he release ,migration and reactor cabin distribution of the noble gas and CsI were analyzed , and the plunge criterion of ventilation system was researched .The results show that in order to assure the radiation dose of the cabin adjacent reactor cabin in the dose limits , the whole ship ventilation system should be plunged within 10 min after the accident . Otherwise ,the protective measure should be used before the whole body dose and thyroid dose reach the dose limits .%本文以严重事故分析程序M ELCOR为计算工具,建立了某型船用堆的计算模型,研究了某型船用堆发生冷段双端断裂大破口失水事故的源项行为及放射性后果。分析了惰性气体Xe与挥发性气体CsI的释放、迁移和舱室分布规律,并对通风系统投入时机进行研究。结果表明:为保证堆舱临舱的剂量辐射在剂量限值内,应于事故发生后10 min内投入全船通风。否则,应于全身剂量和甲状腺剂量达到剂量限值前及时采取防护措施。

  15. Experiment data report for LOFT nonnuclear test L1-3

    Energy Technology Data Exchange (ETDEWEB)

    Millar, G. M.

    1977-04-01

    Test L1-3 was the third in a series of five nonnuclear isothermal blowdown tests conducted by the Loss of Fluid Test (LOFT) Program. For this test the LOFT Facility was configured to simulate a loss-of-coolant accident in a large pressurized water reactor resulting from a 200 percent double-ended shear break in a cold leg of the primary coolant system. A hydraulic core simulator assembly was installed in place of the nuclear core. The initial conditions in the primary coolant system intact loop were: temperature at 540/sup 0/F, pressure at 2256 psig, and loop flow at 2.34 x 10/sup 6/ lbm/hr. During system depressurization, emergency core cooling water was specified to be injected into the lower plenum of the reactor vessel using an accumulator, a low-pressure injection system pump, and a high-pressure injection system pump to provide data on the effects of emergency core cooling on the system thermal-hydraulic response. Injection into the lower plenum was initiated from the high- and low-pressure injection systems. Injection from the accumulator, however, was not initiated because a valve was inadvertently left closed. The experiment, therefore, was not completely successful in that one of the objectives outlined in the experiment operating specification for this test was not accomplished. Test L1-3 was repeated at Test L1-3A to meet the experimental requirements. Despite these difficulties, Test L1-3 did provide very valuable data to verify experiment repeatability.

  16. Experiment data report for semiscale Mod-1 Tests S-03-1, S-03-2, S-03-3, and S-03-4 (reflood heat transfer tests)

    Energy Technology Data Exchange (ETDEWEB)

    Crapo, H.S.; Jensen, M.F.; Sackett, K.E.

    1976-05-01

    Recorded test data are presented for Tests S-03-1, S-03-2, S-03-3, and S-03-4 of the Semiscale Mod-1 reflood heat transfer series (Test Series 3). The tests conducted in this series are separate effects core reflood tests performed to determine the reflood heat transfer characteristics of the 5.5-foot-long Mod-1 rod bundle. Tests S-03-1 through S-03-4 were forced feed reflood tests in which the reflood rate was held constant during each test, although the reflood rate varied from test to test. Tests S-03-1 through S-03-4 were conducted from an initial system temperature of about 290/sup 0/F at a pressure of 60 psia. The electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core during reflood. In all four tests, reflood coolant at 153/sup 0/F was injected directly into the core barrel by means of a specially designed core inlet manifold. Test S-03-4 was performed with a peaked radial power profile, in contrast to the flat profile used in the first three tests. During reflood, core power was reduced from the initial level of 0.097 MW, according to the American Nuclear Society decay heat curve for pressurized water reactor (PWR) core decay heat, +20 percent. The cold leg broken loop piping was open to the pressure suppression system (PSS). A separate steam supply system connected to the PSS system was controlled to maintain constant pressure during the tests.

  17. Response characteristics of HPR1000 primary circuit under different working conditions of the atmospheric relief system after SBLOCA

    Energy Technology Data Exchange (ETDEWEB)

    Sui, Danting, E-mail: suidanting@163.com [School of Nuclear Science and Engineering, North China Electric Power University, Beijing (China); Beijing Key Laboratory of Passive Safety Technology for Nuclear Energy, North China Electric Power University, Beijing (China); Lu, Daogang [School of Nuclear Science and Engineering, North China Electric Power University, Beijing (China); Beijing Key Laboratory of Passive Safety Technology for Nuclear Energy, North China Electric Power University, Beijing (China); Shang, Changzhong; Wei, Yuanyuan [China Nuclear Power Design Co., ltd (ShenZhen), Shenzhen (China); Zhang, Xianjie [School of Nuclear Science and Engineering, North China Electric Power University, Beijing (China); Beijing Key Laboratory of Passive Safety Technology for Nuclear Energy, North China Electric Power University, Beijing (China)

    2017-04-01

    Highlights: • Response of HPR1000 under different VDA conditions after SBLOCA was investigated. • Activation of VDA can trigger ACCU SI earlier with a critical point exists. • VDA capability design should compromise the critical point with reactivity feedback. - Abstract: To cope with SBLOCA in absence of High-Head Safety Injection (HHSI) from design of HPR1000, atmospheric relief system (originally named as VDA in French) is uniquely designed to help to trigger Middle Head Safety Injection (MHSI) or Low Head Safety Injection (LHSI) earlier through cooling primary system quickly after SBLOCA. To make the best use of VDA decay heat removal capability, primary and secondary system of HPR1000 was modeled with RELAP5/SCDAP computer code. After steady-state initialization, a cold leg 30 mm break SBLOCA was simulated with six simulation conditions and five additional cases including availability of ACCU, different VDA discharge locations and area. Response characteristics of primary loop under different VDA working conditions are investigated. Pressurizer pressure decreases rapidly to lower level to trigger the reactor scram, VDA activation and accumulator safety injection sequently. Peak cladding temperature is 899.45 K occurring at 222 s, which is far below the safety limit. Activation of VDA can trigger ACCU SI earlier with a critical point, while positive reactivity will be introduced due to negative moderator temperature effect and Doppler effect. Larger VDA discharge capability will introduce larger reactivity feedback, as well as induce lower core level and SG level. It's suggested that VDA discharge condition should be chosen before the critical point, with the compromise with reactivity feedback introduced due to the negative moderator temperature effect.

  18. Effectiveness of a worksite social & physical environment intervention on need for recovery, physical activity and relaxation; results of a randomized controlled trial.

    Directory of Open Access Journals (Sweden)

    Jennifer K Coffeng

    Full Text Available OBJECTIVE: To investigate the effectiveness of a worksite social and physical environment intervention on need for recovery (i.e., early symptoms of work-related mental and physical fatigue, physical activity and relaxation. Also, the effectiveness of the separate interventions was investigated. METHODS: In this 2 × 2 factorial design study, 412 office employees from a financial service provider participated. Participants were allocated to the combined social and physical intervention, to the social intervention only, to the physical intervention only or to the control group. The primary outcome measure was need for recovery. Secondary outcomes were work-related stress (i.e., exhaustion, detachment and relaxation, small breaks, physical activity (i.e., stair climbing, active commuting, sport activities, light/moderate/vigorous physical activity and sedentary behavior. Outcomes were measured by questionnaires at baseline, 6 and 12 months follow-up. Multilevel analyses were performed to investigate the effects of the three interventions. RESULTS: In all intervention groups, a non-significant reduction was found in need for recovery. In the combined intervention (n = 92, exhaustion and vigorous physical activities decreased significantly, and small breaks at work and active commuting increased significantly compared to the control group. The social intervention (n = 118 showed a significant reduction in exhaustion, sedentary behavior at work and a significant increase in small breaks at work and leisure activities. In the physical intervention (n = 96, stair climbing at work and active commuting significantly increased, and sedentary behavior at work decreased significantly compared to the control group. CONCLUSION: None of the interventions was effective in improving the need for recovery. It is recommended to implement the social and physical intervention among a population with higher baseline values of need for recovery. Furthermore, the

  19. Imprints of a Primordial Preferred Direction on the Microwave Background

    CERN Document Server

    Ackerman, L; Wise, M B; Ackerman, Lotty; Carroll, Sean M.; Wise, Mark B.

    2007-01-01

    Rotational invariance is a well-established feature of low-energy physics. Violations of this symmetry must be extremely small today, but could have been larger in earlier epochs. In this paper we examine the consequences of a small breaking of rotational invariance during the inflationary era when the primordial density fluctuations were generated. Assuming that a fixed-norm vector picked out a preferred direction during the inflationary era, we explore the imprint it would leave on the cosmic microwave background anisotropy, and provide explicit formulas for the expected amplitudes $$ of the spherical-harmonic coefficients. We suggest that it is natural to expect that the imprint on the primordial power spectrum of a preferred spatial direction is approximately scale-invariant, and examine a simple model in which this is true.

  20. Analysis of an accident type sbloca in reactor contention AP1000 with 8.0 Gothic code; Analisis de un accidente tipo Sbloca en la contencion del reactor AP1000 con el codigo Gothic 8.0

    Energy Technology Data Exchange (ETDEWEB)

    Goni, Z.; Jimenez Varas, G.; Fernandez, K.; Queral, C.; Montero, J.

    2016-08-01

    The analysis is based on the simulation of a Small Break Loss-of-Coolant-Accident in the AP1000 nuclear reactor using a Gothic 8.0 tri dimensional model created in the Science and Technology Group of Nuclear Fision Advanced Systems of the UPM. The SBLOCA has been simulated with TRACE 5.0 code. The main purpose of this work is the study of the thermo-hydraulic behaviour of the AP1000 containment during a SBLOCA. The transients simulated reveal close results to the realistic behaviour in case of an accident with similar characteristics. The pressure and temperature evolution enables the identification of the accident phases from the RCS point of view. Compared to the licensing calculations included in the AP1000 Safety Analysis, it has been proved that the average pressure and temperature evolution is similar, yet lower than the licensing calculations. However, the temperature and inventory distribution are significantly heterogeneous. (Author)

  1. Breaking the sound barrier in AdS/CFT

    CERN Document Server

    Hoyos, Carlos; Fernández, David Rodríguez; Vuorinen, Aleksi

    2016-01-01

    It has been conjectured that the speed of sound in holographic models with UV fixed points has an upper bound set by the value of the quantity in conformal field theory. If true, this would set stringent constraints for the presence of strongly coupled quark matter in the cores of physical neutron stars, as the existence of two-solar-mass stars appears to demand a very stiff Equation of State. In this article, we present a family of counterexamples to the speed of sound conjecture, consisting of strongly coupled theories at finite density. The theories we consider include ${\\cal N}=4$ super Yang-Mills at finite R-charge density and non-zero gaugino masses, while the holographic duals are Einstein-Maxwell theories with a minimally coupled scalar in a charged black hole geometry. We show that for a small breaking of conformal invariance, the speed of sound approaches the conformal value from above at large chemical potentials.

  2. A comparison of the effect of the first and second upwind schemes on the predictions of the modified RELAP5/MOD3

    Energy Technology Data Exchange (ETDEWEB)

    Analytis, G.Th. [Paul Scherrer Institute (PSI), Villigen (Switzerland)

    1995-09-01

    As is well-known, both TRAC-BF1 and TRAC-PF are using the first upwind scheme when finite-differencing the phasic momentum equations. In contrast, RELAP5 uses the second upwind which is less diffusive. In this work, we shall assess the differences between the two schemes with our modified version of RELAP5/MOD3 by analyzing some transients of interest. These will include the LOFT LP-LB-1 and LOBI small break LOCA (SB-LOCA) BL34 tests, and a commercial PWR 200% hypothetical large break LOCA (LB-LOCA). In particular, we shall show that for some of these transients, the employment of the first upwind scheme results in significantly different code predictions than the ones obtained when the second upwind scheme is used.

  3. PACTEL: Experiments on the behaviour of the new horizontal steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Kouhia, J.; Riikonen, V.; Purhonen, H. [VTT Energy, Lappeenranta (Finland)

    1995-12-31

    Experiments were performed to study the behaviour of the PACTEL facility, a medium scale integral test loop simulating VVER 440 pressurized water reactors. The study focused on the operation of the new horizontal steam generator model installed in PACTEL. Three experiments were conducted: a small-break test to observe the steam generator behaviour over a range of primary coolant inventories, a hot leg loop seal experiment to study the cyclic behaviour of a loop seal and a loss of secondary side feedwater test to examine the effect of uncovered tubes in the steam generator. A reverse flow was observed in the lower part of the U-tube bundle of the steam generator during natural circulation. The flow reversal point dropped when the tubes uncovered, during secondary inventory reduction. (orig.). 5 refs.

  4. System approach in the investigation of coolant parametrical oscillations in passive safety injection systems (PSIS)

    Energy Technology Data Exchange (ETDEWEB)

    Proskouriakov, K.N. [Moskovskij Ehnergeticheskij Inst., Moscow (Russian Federation)

    2001-07-01

    The use of thermal-hydraulic computer codes is an important part of the work programme for activities in the field of nuclear power plants (NPP) Safety Research as it will enable to define better the test configuration and parameter range extensions and to extrapolate the results of the small scale experiments towards full scale reactor applications. The CATHARE2, RELAP5, the WCOBRA/TRAC, and APROS codes are the estimate thermal hydraulic codes for the evaluation of large and small break loss of coolant accidents (LOCA). The relatively good agreement experimental data with the calculations have been presented. There was shown also some big mistakes in predicting distribution of flow when two phase are present. Model of parametrical oscillation (P.O.) worked out gives explanation for flow oscillations and indicates that the phenomenon of P.O. appears under certain combination of thermal-hydraulic parameters and structure of heat-removal system. (orig.)

  5. Scram discharge volume break studies accident sequence analysis

    Energy Technology Data Exchange (ETDEWEB)

    Harrington, R.M.; Hodge, S.A.

    1982-01-01

    This paper is a summary of a report describing the predicted response of Unit 1 at the Tennessee Valley Authority (TVA) Browns Ferry Nuclear Plant to a hypothetical small break loss of coolant accident (SBLOCA) outside of containment. The accident studied would be initiated by a break in the scram discharge volume (SDV) piping when it is pressurized to full reactor vessel pressure as a normal consequence of a reactor scram. If the scram could be reset, the scram outlet valves would close to isolate the SDV and the piping break from the reactor vessel. However, reset is possible only if the conditions that caused the scram have cleared; it has been assumed in this study that the scram signal remains in effect over a long period of time.

  6. Viability of bi-maximal solution of the Zee mass matrix

    CERN Document Server

    Brahmachari, B; Brahmachari, Biswajoy; Choubey, Sandhya

    2002-01-01

    We know $L_e-L_\\mu-L_\\tau$ symmetry gives $m^2_1= m^2_2 >> m^2_3$ pattern in Zee model. $\\Delta m^2_\\odot$ emerges from a small breaking of this symmetry. Because this symmetry is broken very weakly $\\theta_\\odot$ does not deviate much from $\\tan^2 \\theta_\\odot=1$ which is its value in the symmetric limit. This gives a mismatch with LMA solution where mixing is large but not exactly maximal. We confront this property of Zee mass matrix by phenomenologically analyzing recent results from solar and atmospheric neutrino oscillation experiments at various confidence levels. We conclude that LOW type solution is compatible with the Zee mass matrix at 99% confidence level when atmospheric neutrino deficit is explained by maximal $\

  7. Simulation of research loop LOBI-MOD2 with RELAP5/MOD3.3 code for LOBI thermo hydraulic test A1-93

    Energy Technology Data Exchange (ETDEWEB)

    Pesaran, Farshad; Barati, Ramin [Islamic Azad Univ., Shiraz (Iran, Islamic Republic of). Dept. of Electrical Engineering

    2016-06-15

    RELAP5/MOD3.3 is one of the used computer codes for the simulation of event thermal-hydraulics of nuclear power plants. The LOBI test facility is a full-power high-pressure integral system test facility, representing an approximately 1: 700 scale model of a 4-loop, 1300 MWe PWR. A new simulation of the small break LOCA test A1-93 has been carried out in a LOBI/Mod2 facility for reaching good agreement and to evaluate the performance of the RELAP5/MOD3.3 code. Good agreement was obtained in general between the code predictions and the experimental data in transient state.

  8. Analytical and experimental research into boron dilution events

    Energy Technology Data Exchange (ETDEWEB)

    Teschendorff, V. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany); Umminger, K. [Framatome ANP GmbH, Erlengen (Germany); Weiss, F.P. [Forschungszentrum Rossendorf e.V. (FZR) (Germany). Inst. fuer Sicherheitsforschung

    2001-07-01

    Research activities are being performed in Germany with the aim to improve and validate the methods for predicting boron dilution events. Integral experiments in the PKL test facility investigate the thermal-hydraulic system behaviour in a wide range of conditions. The latest test program comprises small break LOCA scenarios with boron dilution. For these tests, boric acid in the coolant is used together with an advanced instrumentation that can measure boron concentration during the transient. Mixing processes in the downcomer and lower plenum under the influence of various loop operating conditions are studied in the transparent 1:5 ROCOM four-loop test facility equipped with advanced wire mesh sensors to follow the transient concentration patterns. Analytical R and D activities include further model development and validation in the thermal-hydraulic system code ATHLET as well as assessment calculations for detailed three-dimensional mixing in the reactor pressure vessel with CFD-codes. (authors)

  9. Transient two-phase performance of LOFT reactor coolant pumps

    Energy Technology Data Exchange (ETDEWEB)

    Chen, T.H.; Modro, S.M.

    1983-01-01

    Performance characteristics of Loss-of-Fluid Test (LOFT) reactor coolant pumps under transient two-phase flow conditions were obtained based on the analysis of two large and small break loss-of-coolant experiments conducted at the LOFT facility. Emphasis is placed on the evaluation of the transient two-phase flow effects on the LOFT reactor coolant pump performance during the first quadrant operation. The measured pump characteristics are presented as functions of pump void fraction which was determined based on the measured density. The calculated pump characteristics such as pump head, torque (or hydraulic torque), and efficiency are also determined as functions of pump void fractions. The importance of accurate modeling of the reactor coolant pump performance under two-phase conditions is addressed. The analytical pump model, currently used in most reactor analysis codes to predict transient two-phase pump behavior, is assessed.

  10. Analysis of human actions at SBLOCA in PWR sequences with loss and recovery of HPSI; Analisis de las actuaciones humanas en secuencias de SBLOCA en PWR con perdida y recuperacion de HPSI

    Energy Technology Data Exchange (ETDEWEB)

    Montero, J.; Gonzalez-Cadelo, J.; Queral, C.

    2014-10-01

    Emergency Operating Procedures (EOP) of US PWRs establish that reactor coolant pumps (RCP) should be tripped during a Small-Break Loss of Coolant Accident (SBLOCA) by the operating crew, provided that the subcooling margin has been lost at the core outlet and the High-Pressure Safety Injection (HPSI) is available. On the other hand, it HPSI is unavailable, RCPs must remain in operation. In this work, it is analyzed human actions in SBLOCA sequences with PHSI failure (and therefore without RCP trip), but with a subsequent HPSI recovery (and therefore with a subsequent RCP trip). The analysis was performed with the TRACE code by means of a model with conservative assumptions. The results show that the HPSI recovery and the subsequent RCP trip can lead the plant to damage conditions it recovery occur when the vessel level is lower. Results also show that such damage can be avoided if only 2 out of 3 RCPs are tripped. (Author)

  11. Breaking the sound barrier in holography

    Science.gov (United States)

    Hoyos, Carlos; Jokela, Niko; Rodríguez Fernández, David; Vuorinen, Aleksi

    2016-11-01

    It has been conjectured that the speed of sound in holographic models with UV fixed points has an upper bound set by the value of the quantity in conformal field theory. If true, this would set stringent constraints for the presence of strongly coupled quark matter in the cores of physical neutron stars, as the existence of two-solar-mass stars appears to demand a very stiff equation of state. In this article, we present a family of counterexamples to the speed of sound conjecture, consisting of strongly coupled theories at finite density. The theories we consider include N =4 super Yang-Mills at finite R -charge density and nonzero gaugino masses, while the holographic duals are Einstein-Maxwell theories with a minimally coupled scalar in a charged black hole geometry. We show that for a small breaking of conformal invariance, the speed of sound approaches the conformal value from above at large chemical potentials.

  12. Methods development to evaluate the risk of upgrading to DCS: The human factor

    Energy Technology Data Exchange (ETDEWEB)

    Ostrom, L.T.; Wilhelmsen, C.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1995-04-01

    The NRC recognizes that a more complete technical basis for understanding and regulating advanced digital technologies in commercial nuclear power plants is needed. A concern is that the introduction of digital safety systems may have an impact on risk. There is currently no standard methodology for measuring digital system reliability. A tool currently used to evaluate NPP risk in analog systems is the probabilistic risk assessment (PRA). The use of this tool to evaluate the digital system risk was considered to be a potential methodology for determining the risk. To test this hypothesis, it was decided to perform a limited PRA on a single dominant accident sequence. However, a review of existing human reliability analysis (HRA) methods showed that they were inadequate to analyze systems utilizing digital technology. A four step process was used to adapt existing HRA methodologies to digital environments and to develop new techniques. The HRA methods were then used to analyze an NPP that had undergone a backfit to digital technology in order to determine, as a first step, whether the methods were effective. The very small-break loss of coolant accident sequence was analyzed to determine whether the upgrade to the Eagle-21 process protection system had an effect on risk. The analysis of the very small-break LOCA documented in the Sequoyah PRA was used as the basis of the analysis. The analysis of the results of the HRA showed that the mean human error probabilities for the Eagle-21 PPS were slightly less than those for the analog system it replaced. One important observation from the analysis is that the operators have increased confidence steming from the better level of control provided by the digital system. The analysis of the PRA results, which included the human error component and the Eagle-21 PPS, disclosed that the reactor protection system had a higher failure rate than the analog system, although the difference was not statistically significant.

  13. Loss of Coolant Accident Analysis Methodology for SMART-P

    Energy Technology Data Exchange (ETDEWEB)

    Bae, K. H.; Lee, G. H.; Yang, S. H.; Yoon, H. Y.; Kim, S. H.; Kim, H. C

    2006-02-15

    The analysis methodology on the Loss-of-coolant accidents (LOCA's) for SMART-P is described in this report. SMART-P is an advanced integral type PWR producing a maximum thermal power of 65.5 MW with metallic fuel. LOCA's are hypothetical accidents that would result from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant makeup system, from breaks in pipes in the reactor coolant pressure boundary up to and including a break equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant system. Since SMART-P contains the major primary circuit components in a single Reactor Pressure Vessel (RPV), the possibility of a large break LOCA (LBLOCA) is inherently eliminated and only the small break LOCA is postulated. This report describes the outline and acceptance criteria of small break LOCA (SBLOCA) for SMART-P and documents the conservative analytical model and method and the analysis results using the TASS/SMR code. This analysis method is applied in the SBLOCA analysis performed for the ECCS performance evaluation which is described in the section 6.3.3 of the safety analysis report. The prediction results of SBLOCA analysis model of SMART-P for the break flow, system's pressure and temperature distributions, reactor coolant distribution, single and two-phase natural circulation phenomena, and the time of major sequence of events, etc. should be compared and verified with the applicable separate and integral effects test results. Also, it is required to set-up the feasible acceptance criteria applicable to the metallic fueled integral reactor of SMART-P. The analysis methodology for the SBLOCA described in this report will be further developed and validated as the design and licensing status of SMART-P evolves.

  14. Study on Design of 600 MW PWR Accumulator%600 MW 压水堆安注箱设计研究

    Institute of Scientific and Technical Information of China (English)

    冯进军; 冯文卿; 周克峰; 杨志义; 石俊英; 种毅敏; 柴国旱

    2015-01-01

    In this paper ,the TRACE and SNAP were used to establish two‐loop PWR thermal hydraulic system analysis model . The different accumulator design schemes were calculated and analyzed under LBLOCA .The safety injection effect was accessed according to simulation results by comparing peak cladding temperature of each design under LBLOCA .In the end ,the possible way to optimize design was found through this study .The research results show that the upper plenum and downcomer injection at the same time is more effective than the cold leg injection or the downcomer injection ,and the proper selection of initial accumulator pressure can lower peak cladding temperature and increase LOCA safety margin .%本文用美国核管会热工水力程序 TRACE 和图形化建模软件 SNAP ,建立了600 MW 两环路压水堆一回路和二回路热工水力系统分析模型,并对安注箱的各设计方案进行大破口失水事故(LBLOCA)模拟计算,通过对比各设计方案在 LBLOCA 事故下计算出的峰值包壳温度,研究安注箱在大破口失水事故工况下的安注性能,最后给出了优化的设计方案,并提出了可行的设计改进建议。研究结果表明,上腔室和下降段同时注入的方式较冷段注入和下降段注入更有效,且恰当地选取初始安注箱压力,可有效降低峰值包壳温度,提高 LOCA 裕量。

  15. Recent safety issues concerning steam generators in France and their analysis by IRSN

    Energy Technology Data Exchange (ETDEWEB)

    Sollier, T.; Le Calvar, M.; Balestreri, F.; Mermaz, F. [Inst. de Radioprotection ed de Surete Nucleaaire (IRSN) (France)

    2009-07-01

    In France between 2004 and 2008, there were recurrent safety issues concerning the operation of Steam Generators (SGs). Among these issues, at least three are generic to the EDF Nuclear Power Plant (NPP) fleet: In 2004, 2005 and 2006, a total of three primary to secondary leaks occurred at Cruas NPP. The root cause of these leaks was a modification of the thermal-hydraulic condition of the SG due to a heavy build-up of oxide deposits at the flow holes of the quatrefoil-shaped Tube Support Plates (TSPs). The clogging of the TSPs, meant that the water/steam flow accelerated at the U-bend location and that tubes were subjected to high cycle fatigue near the uppermost TSPs due to flow-induced vibration. For each unscheduled outage, the origin of the leaks was a circumferential fatigue crack located at the upper edge of the uppermost TSP; In 2008, a primary to secondary leak occurred at Fessenheim NPP. The source of the leak was a circumferential crack located at the edge of the uppermost TSP at approximately the same location where cracks were found on Cruas Units. However, the SGs of Fessenheim Unit 2 have circular flow holes without significant flow section reduction due to oxide deposits. The root cause of the event was determined to be fluid-elastic instability in the U-Bend for a tube not supported by an Anti-Vibration Bar (AVB). The AVB position in the tube bundle deviated from the manufacturing design, something which affects a large number of SGs in France; In 2008, a plug failure was observed at Saint Alban NPP. A plug was propelled from the hot to the cold leg during the primary coolant circuit hydrotest. The plugging operation had been performed before the hydrotest. In this paper, IRSN presents its technical analysis of these events. It includes the SG secondary side water conditioning operation, the non-destructive testing methods in relation to the clogging-rate evaluation and tube integrity assessment, and the mechanical issues due to tube vibration

  16. An integrated model of tritium transport and corrosion in Fluoride Salt-Cooled High-Temperature Reactors (FHRs) – Part I: Theory and benchmarking

    Energy Technology Data Exchange (ETDEWEB)

    Stempien, John D., E-mail: john.stempien@inl.gov; Ballinger, Ronald G., E-mail: hvymet@mit.edu; Forsberg, Charles W., E-mail: cforsber@mit.edu

    2016-12-15

    transfer of corrosion products from the hot to the cold leg (as was observed in the experiments with salts containing UF{sub 4}). In a separate paper the results of TRIDENT simulations in a prototypical FHR are presented.

  17. Tracking the position of the underwater robot for nuclear reactor inspection

    Energy Technology Data Exchange (ETDEWEB)

    Jeo, J. W.; Kim, C. H.; Seo, Y. C.; Choi, Y. S.; Kim, S. H. [KAERI, Taejon (Korea, Republic of)

    2003-10-01

    The tracking procedure of the underwater mobile robot moving and submerging ahead to nuclear reactor vessel for visual inspection, which is required to find the foreign objects such as loose parts, is described. The yellowish underwater robot body tends to present a big contrast to boron solute cold water of nuclear reactor vessel, tinged with indigo by the Cerenkov effect. In this paper, we have found and tracked the positions of underwater mobile robot using the two color information, yellow and indigo. From the horizontal and vertical profiles analysis of the color image, the blue, green, and the gray component have the inferior signal-to-noise characteristics compared to the red component. The center coordinates extraction procedures areas follows. The first step is to segment the underwater robot body to cold water with indigo background. From the RGB color components of the entire monitoring image taken with the color CCD camera, we have selected the red color component. In the selected red image, we extracted the positions of the underwater mobile robot using the following process sequences; binarization, labelling, and centroid extraction techniques. In the experiment carried out at the Youngkwang unit 5 nuclear reactor vessel, we have tracked the center positions of the underwater robot submerged near the cold leg and the hot leg way, which is fathomed to 10m deep in depth. When the position of the robot vehicle fluctuates between the previous and the current image frame due to the flickering noise and light source, installed temporally in the bottom of the reactor vessel, we adaptively adjusted the ROI window. Adding the ROI windows of the previous frame to the current frame, and then setting up the ROI window of the next image frame, we can robustly track the positions of the underwater robot and control the target position's divergence. From these facts, we can conclude that using the red component from color camera is more efficient tracking

  18. Westinghouse Small Modular Reactor nuclear steam supply system design

    Energy Technology Data Exchange (ETDEWEB)

    Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

  19. Numerical simulation of vapor-liquid two-phase flow and boiling in steam generator based on population balance%基于群体平衡原理的蒸汽发生器汽液两相流动与沸腾数值模拟

    Institute of Scientific and Technical Information of China (English)

    杨元龙; 孙宝芝; 杨柳; 郑陆松

    2014-01-01

    Based on similarity principle, a three-dimensional“unit pipe”physical model is built for a steam generator coupled with quatrefoil tube support plates. A multiple size group model considering breakup and coalescence of bubbles is used to describe bubble size distribution and hydraulic characteristic in the secondary side. The process of local vapor-liquid two-phase boiling with phase change is calculated through a thermal phase change model. Numerical investigation is carried out on vapor-liquid two-phase flow and boiling of steam generator at Daya Bay Nuclear Power Plant. Simulation results show that vapor and liquid velocities rapidly increase in the support plate and jet flow appears, and when the flow leaves the support plate orifice, reflow forms rapidly. The periodic distribution from small to large bubble size occurs obviously within adjacent tube support plates and the maximum bubble diameter decreases slowly in the direction of hot and cold legs. In addition, the averaged heat transfer coefficient of secondary side agrees well with the results calculated using Rohsenow’s correlation.%基于相似原理建立了耦合四叶梅花形支撑板的蒸汽发生器“单元管”三维物理模型。采用考虑汽泡聚合与破碎效应的MUSIG (multiple-size-group)模型描述二次侧汽泡尺度分布和水力特性,热相变模型计算二次侧汽液两相沸腾相变过程,对大亚湾蒸汽发生器汽液两相流动与沸腾过程进行数值研究。模拟结果表明,支撑板位置处汽、液相流速均急剧升高,产生射流,在离开支撑板孔口时迅速形成回流。两相邻支撑板间出现明显的汽泡由小到大的周期性变化过程,冷、热端沿程汽泡最大直径缓慢减小。二次侧平均传热系数与 Rohsenow 经验关联式的计算结果吻合较好。

  20. RELAP5/MOD3.2 post-test analysis and CIAU uncertainty evaluation of LOFT experiment L2-5

    Energy Technology Data Exchange (ETDEWEB)

    Alessandro Petruzzi; Walter Giannotti [DIMNP, Universit y of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy); Francesco D' Auria [DIMNP, Universit y of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)

    2005-07-01

    Full text of publication follows: The paper deals with the activity performed at University of Pisa in the framework of the participation to the Phase II and III of the BEMUSE (Best Estimate Methods - Uncertainty and Sensitivity Evaluation) Programme. The scope of the Programme is to perform Large Break Loss-Of-Coolant Accident (LBLOCA) analyses making reference to experimental data and to a Nuclear Power Plant (NPP) in order to address the issue of 'the capabilities of computational tools' including scaling/uncertainty analysis. The justification for such an activity comes from the consideration that a wide spectrum of uncertainty methods applied to Best Estimate codes exist and are used in research laboratories, but their practicability and/or validity is not sufficiently established to support general use of the codes and acceptance by industry and safety authorities. The consideration of the Best Estimate codes and uncertainty evaluation for Design Basis Accident (DBA), by itself, shows the safety significance of the proposed activity.The Phase II of BEMUSE Programme is connected with the reanalysis of the Experiment L2 -5 performed in the LOFT (Loss Of Fluid Test) facility in June 1982. The LOFT facility, installed at the Idaho National Engineering Laboratory (INEL), is a 50 MWth Pressurized Water Reactor (PWR) system with instruments that measure and provide data on the system thermal-hydraulic and nuclear conditions. The operation of the LOFT is typical of large ({approx}1000 MWe) commercial PWR. For the performance of Experiment L2-5, the LOFT facility was configured to simulate a double-ended 200 % cold leg break in a four-loop PWR operating at nominal conditions. Assumption of loss of offsite power and atypical primary coolant pump coast down were incorporated into the simulation to create core flow stagnation. The light water reactor transient analysis code Relap5/Mod3.2 has been used to simulate this experiment and the standard procedure

  1. Energy Conversion Advanced Heat Transport Loop and Power Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Oh, C. H.

    2006-08-01

    operating conditions as well as trade offs between efficiency and capital cost. Prametric studies were carried out on reactor outlet temperature, mass flow, pressure, and turbine cooling. Recommendations on the optimal working fluid for each configuration were made. A steady state model comparison was made with a Closed Brayton Cycle (CBC) power conversion system developed at Sandia National Laboratory (SNL). A preliminary model of the CBC was developed in HYSYS for comparison. Temperature and pressure ratio curves for the Capstone turbine and compressor developed at SNL were implemented into the HYSYS model. A comparison between the HYSYS model and SNL loop demonstrated power output predicted by HYSYS was much larger than that in the experiment. This was due to a lack of a model for the electrical alternator which was used to measure the power from the SNL loop. Further comparisons of the HYSYS model and the CBC data are recommended. Engineering analyses were performed for several configurations of the intermediate heat transport loop that transfers heat from the nuclear reactor to the hydrogen production plant. The analyses evaluated parallel and concentric piping arrangements and two different working fluids, including helium and a liquid salt. The thermal-hydraulic analyses determined the size and insulation requirements for the hot and cold leg pipes in the different configurations. Economic analyses were performed to estimate the cost of the va

  2. SBLOCA AND LOFW EXPERIMENTS IN A SCALED-DOWN IET FACILITY OF REX-10 REACTOR

    Directory of Open Access Journals (Sweden)

    YEON-GUN LEE

    2013-06-01

    Full Text Available This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA and the loss-of-feedwater accident (LOFW in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF, a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

  3. Unified flavor symmetry from warped dimensions

    Directory of Open Access Journals (Sweden)

    Mariana Frank

    2015-03-01

    Full Text Available In a model of warped extra-dimensions with all matter fields in the bulk, we propose a scenario which explains all the masses and mixings of the SM fermions. In this scenario, the same flavor symmetric structure is imposed on all the fermions of the Standard Model (SM, including neutrinos. Due to the exponential sensitivity on bulk fermion masses, a small breaking of this symmetry can be greatly enhanced and produce seemingly un-symmetric hierarchical masses and small mixing angles among the charged fermion zero-modes (SM quarks and charged leptons, thus washing out visible effects of the symmetry. If the Dirac neutrinos are sufficiently localized towards the UV boundary, and the Higgs field leaking into the bulk, the neutrino mass hierarchy and flavor structure will still be largely dominated and reflect the fundamental flavor structure, whereas localization of the quark sector would reflect the effects of the flavor symmetry breaking sector. We explore these features in an example based on which a family permutation symmetry is imposed in both quark and lepton sectors.

  4. A containment analysis for SBLOCA without ECI in the refurbished Wolsong-1 Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T.M.; Moon, B.J.; Bae, C.J.; Lee, S.H.; Choi, C.J.; Lee, D.S. [NSSS, Korea Power Engineering Company, Inc., Daejeon (Korea, Republic of); Kim, S.M. [NETEC, Korea Hydro and Nuclear Power Company, Inc., Daejeon (Korea, Republic of)

    2010-07-01

    A small break leading to loss of coolant accident (SBLOCA), being one of the topic accidents in the nuclear plant diagnosis in recent years, has been analyzed and evaluated for the refurbished Wolsong-1 Nuclear Power Plant (NPP). The industry standard toolset (IST) codes developed by CANDU Owners Group and updated models including design change parameters are applied to the event analyses. GOTHIC code has been used for the containment analysis of Wolsong-1. Also, SMART-IST code fitted in the Iodine Chemistry (IMOD-2) model has been used to predict nuclide behavior within the containment considering various aspects. IMOD-2 was incorporated into SMART-IST as a module dealing the chemical transformations and mass transfer of iodine species in containment. IMOD-2 model is very sensitive to paint and chemicals. The parameter studies for IMOD-2 model are performed to decide the analysis value set. The developed methodology and the results of SBLOCA without ECI are presented herein. Under the most heat-up conditions, the radionuclide release from the failed fuel into the containment and subsequently to the environment is such that the radioactive doses to the public are below the acceptable limits. (author)

  5. Animation model of Krsko nuclear power plant for RELAP5 calculations

    Energy Technology Data Exchange (ETDEWEB)

    Prosek, Andrej, E-mail: andrej.prosek@ijs.s [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Mavko, Borut [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

    2011-04-15

    Today most software applications, also in the nuclear field, come with a graphical user interface. The first graphical user interface for the RELAP5 thermal-hydraulic computer code was called the Nuclear Plant Analyzer (NPA). Later, Symbolic Nuclear Analysis Package (SNAP) was developed. The purpose of the present study was to develop SNAP animation model of Krsko nuclear power plant (NPP) for RELAP5 calculations with the aim to help analyze the results. In addition, the reference calculations for Krsko full scope simulator validation were performed with the latest RELAP5/MOD3.3 Patch 03 code and compared to previous RELAP5 versions to provide verified source data, needed to demonstrate animation model. In total six scenarios were analyzed: two scenarios of the small-break loss-of-coolant accident, two scenarios of the loss of main feedwater, a scenario of the anticipated transient without scram, and a scenario of the steam generator tube rupture. The use of SNAP for animation of Krsko nuclear power plant analyses showed several benefits, especially better understanding of the calculated physical phenomena and processes. It can be concluded that an animation tool was created, which enables to analyze very complex accident scenarios. The graphical surface helps keeping the overview and focusing on the main influences. Also, the use of such support tools to system codes may significantly contribute to better quality of safety analysis.

  6. Water flow based geometric active deformable model for road network

    Science.gov (United States)

    Leninisha, Shanmugam; Vani, Kaliaperumal

    2015-04-01

    A width and color based geometric active deformable model is proposed for road network extraction from remote sensing images with minimal human interception. Orientation and width of road are computed from a single manual seed point, from which the propagation starts both right and left hand directions of the starting point, which extracts the interconnected road network from the aerial or high spatial resolution satellite image automatically. Here the propagation (like water flow in canal with defined boundary) is restricted with color and width of the road. Road extraction is done for linear, curvilinear (U shape and S shape) roads first, irrespective of width and color. Then, this algorithm is improved to extract road with junctions in a shape of L, T and X along with center line. Roads with small break or disconnected roads are also extracts by a modified version of this same algorithm. This methodology is tested and evaluated with various remote sensing images. The experimental results show that the proposed method is efficient and extracting roads accurately with less computation time. However, in complex urban areas, the identification accuracy declines due to the various sizes of obstacles, over bridges, multilane etc.

  7. Simulation of the Westinghouse AP1000 Response to SBLOCA Using RELAP/SCDAPSIM

    Directory of Open Access Journals (Sweden)

    Ayah Elshahat

    2014-01-01

    Full Text Available Great interest is given now to advanced nuclear reactors especially those using passive safety components. The Westinghouse AP1000 Advanced Passive pressurized water reactor (PWR is an 1117 MWe PWR designed to achieve a high safety and performance record. The AP1000 safety system uses natural driving forces, such as pressurized gas, gravity flow, natural circulation flow, and convection. In this paper, the safety performance of the AP1000 during a small break loss of coolant accident (SBLOCA is investigated. This was done by modelling the AP1000 and the passive safety systems employed using RELAP/SCDAPSIM code. RELAP/SCDAPSIM is designed to describe the overall reactor coolant system (RCS thermal hydraulic response and core behaviour under normal operating conditions or under design basis or severe accident conditions. Passive safety components in the AP1000 showed a clear improvement in accident mitigation. It was found that RELAP/SCDAPSIM is capable of modelling a LOCA in an AP1000 and it enables the investigation of each safety system component response separately during the accident. The model is also capable of simulating natural circulation and other relevant phenomena. The results of the model were compared to that of the NOTRUMP code and found to be in a good agreement.

  8. Reconstructing the past: methods and techniques for the digital restoration of fossils

    Science.gov (United States)

    Lautenschlager, Stephan

    2016-10-01

    During fossilization, the remains of extinct organisms are subjected to taphonomic and diagenetic processes. As a result, fossils show a variety of preservational artefacts, which can range from small breaks and cracks, disarticulation and fragmentation, to the loss and deformation of skeletal structures and other hard parts. Such artefacts can present a considerable problem, as the preserved morphology of fossils often forms the basis for palaeontological research. Phylogenetic and taxonomic studies, inferences on appearance, ecology and behaviour and functional analyses of fossil organisms strongly rely on morphological information. As a consequence, the restoration of fossil morphology is often a necessary prerequisite for further analyses. Facilitated by recent computational advances, virtual reconstruction and restoration techniques offer versatile tools to restore the original morphology of fossils. Different methodological steps and approaches, as well as software are outlined and reviewed here, and advantages and disadvantages are discussed. Although the complexity of the restorative processes can introduce a degree of interpretation, digitally restored fossils can provide useful morphological information and can be used to obtain functional estimates. Additionally, the digital nature of the restored models can open up possibilities for education and outreach and further research.

  9. Experimental Studies for the VVER-440/213 Bubble Condenser System for Kola NPP at the Integral Test Facility BC V-213

    Directory of Open Access Journals (Sweden)

    Vladimir N. Blinkov

    2012-01-01

    Full Text Available In the frame of Tacis Project R2.01/99, which was running from 2003 to 2005, the bubble condenser system of Kola NPP (unit 3 was qualified at the integral test facility BC V-213. Three LB LOCA tests, two MSLB tests, and one SB LOCA test were performed. The appropriate test scenarios for BC V-213 test facility, modeling accidents in the Kola NPP unit 3, were determined with pretest calculations. Analysis of test results has shown that calculated initial conditions and test scenarios were properly reproduced in the tests. The detailed posttest analysis of the tests performed at BC V-213 test facility was aimed to validate the COCOSYS code for the calculation of thermohydraulic processes in the hermetic compartments and bubble condenser. After that the validated COCOSYS code was applied to NPP calculations for Kola NPP (unit 3. Results of Tacis R2.01/99 Project confirmed the bubble condenser functionality during large and small break LOCAs and MSLB accidents. Maximum loads were reached in the LB LOCA case. No condensation oscillations were observed.

  10. Experimental Study of SBLOCA Simulation of Safety-Injection Line Break with Single Train Passive Safety System of SMART-ITL

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Sung Uk; Bae, Hwang; Ryu, Hyo Bong; Byun, Sun Joon; Kim, Woo Shik; Shin, Yong Cheol; Yi, Sung Jae; Park, Hyun Sik [KAERI, Daejon (Korea, Republic of)

    2016-03-15

    An experimental study of the thermal-hydraulic characteristics of passive safety systems (PSSs) was conducted using a system-integrated modular advanced reactor-integral test loop (SMART-ITL). The present passive safety injection system for the SMART-ITL consists of one train with the core makeup tank (CMT), the safety injection tank, and the automatic depressurization system. The objective of this study is to investigate the injection effect of the PSS on the small-break loss-of-coolant accident (SBLOCA) scenario for a 0.4 inch line break in the safety-injection system (SIS). The steady-state condition was maintained for 746 seconds before the break. When the major parameters of the target value and test results were compared, most of the thermal-hydraulic parameters agreed closely with each other. The water level of the reactor pressure vessel (RPV) was maintained higher than that of the fuel assembly plate during the transient, for the present CMT and safety injection tank (SIT) flow rate conditions. It can be seen that the capability of an emergency core cooling system is sufficient during the transient with SMART passive SISs.

  11. Integral Test Facility PKL: Experimental PWR Accident Investigation

    Directory of Open Access Journals (Sweden)

    Klaus Umminger

    2012-01-01

    Full Text Available Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circulation pumps and steam generators (SGs arranged symmetrically around the reactor pressure vessel (RPV. The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermal-hydraulic phenomena. This paper presents a survey of test objectives and programs carried out to date. It also describes the test facility in its present state. Some important results obtained over the years with focus on investigations carried out since the beginning of the international cooperation are exemplarily discussed.

  12. SCDAP/RELAP5/MOD 3.1 code manual: MATPRO, A library of materials properties for Light-Water-Reactor accident analysis. Volume 4

    Energy Technology Data Exchange (ETDEWEB)

    Hagrman, D.T. [ed.; Allison, C.M.; Berna, G.A. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)] [and others

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light -- water-reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume, Volume IV, describes the material properties correlations and computer subroutines (MATPRO) used by SCDAP/RELAP5. formulation of the materials properties are generally semi-empirical in nature. The materials property subroutines contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, cadmium, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, fill gas mixtures, carbon steel, and tungsten. This document also contains descriptions of the reaction and solution rate models needed to analyze a reactor accident.

  13. Preliminary accident analysis of Flexblue® underwater reactor

    Directory of Open Access Journals (Sweden)

    Haratyk Geoffrey

    2015-01-01

    Full Text Available Flexblue® is a subsea-based, transportable, small modular reactor delivering 160 MWe. Immersion provides the reactor with an infinite heat sink – the ocean – around the metallic hull. The reference design includes a loop-type PWR with two horizontal steam generators. The safety systems are designed to operate passively; safety functions are fulfilled without operator action and external electrical input. Residual heat is removed through four natural circulation loops: two primary heat exchangers immersed in safety tanks cooled by seawater and two emergency condensers immersed in seawater. In case of a primary piping break, a two-train safety injection system is actuated. Each train includes a core makeup tank, an accumulator and a safety tank at low pressure. To assess the capability of these features to remove residual heat, the reactor and its safety systems have been modelled using thermal-hydraulics code ATHLET with conservative assumptions. The results of simulated transients for three typical PWR accidents are presented: a turbine trip with station blackout, a large break loss of coolant accident and a small break loss of coolant accident. The analyses show that the safety criteria are respected and that the reactor quickly reaches a safe shutdown state without operator action and external power.

  14. Flow measurement by pulsed-neutron activation techniques at the PKL facility at Erlangen (Germany). [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kehler, P.

    1982-03-01

    Flow velocities in the downcomer at the PKL facility (in Erlangen, Germany) were measured by the Pulsed-Neutron Activation (PNA) techniques. This was the first time that a fully automated PNA system, incorporating a dedicated computer for on-line data reduction, was used for flow measurements. A prototype of a portable, pulsed, high-output neutron source, developed by the Sandia National Laboratories for the US Nuclear Regulatory Commission, was also successfully demonstrated during this test. The PNA system was the primary flow-measuring device used at the PKL, covering the whole range of velocities of interest. In this test series, the PKL simulated small-break accidents similar to the one that occurred at TMI. The flow velocities in the downcomer were, therefore, very low, ranging between 0.03 and 0.35 m/sec. Two additional flow-measuring methods were used over a smaller range of velocities. Wherever comparison was possible, the PNA-derived velocity values agreed well with the measurements performed by the two more conventional methods.

  15. Prediction of LOCA Break Size Using CFNN

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Geon Pil; Yoo, Kwae Hwan; Back, Ju Hyun; Kim, Dong Yeong; Na, Man Gyun [Chosun University Gwangju (Korea, Republic of)

    2016-05-15

    The NPPs have the emergency core cooling system (ECCS) such as a safety injection system. The ECCS may not function properly in case of the small break size due to a slight change of pressure in the pipe. If the coolant is not supplied by ECCS, the reactor core will melt. Therefore, the meltdown of reactor core have to be prevented by appropriate accident management through the prediction of LOCA break size in advance. This study presents the prediction of LOCA break size using cascaded fuzzy neural network (CFNN). The CFNN model repeatedly applies FNN modules that are serially connected. The CFNN model is a data-based method that requires data for its development and verification. The data were obtained by numerically simulating severe accident scenarios of the optimized power reactor (OPR1000) using MAAP code, because real severe accident data cannot be obtained from actual NPP accidents. The CFNN model has been designed to rapidly predict the LOCA break size in LOCA situations. The CFNN model was trained by using the training data set and checked by using test data set. These data sets were obtained using MAAP code for OPR1000 reactor. The performance results of the CFNN model show that the RMS error decreases as the stage number of the CFNN model increases. In addition, the performance result of the CFNN model presents that the RMS error level is below 4%.

  16. The noncondensable gas effects on loss-of-coolant accident steam condensation loads in boiling water reactor pressure suppression pool

    Energy Technology Data Exchange (ETDEWEB)

    Kukita, Y.; Namatame, K.; Shiba, M.; Takeshita, I.

    1983-11-01

    The noncondensable gas effects on the loss-ofcoolant-accident-induced steam condensation loads in the boiling water reactor pressure suppression pool have been investigated with regard to experimental data obtained from a large-scale multivent test program. Previous studies have noted that the presence of the noncondensable gas (air), which initially fills the containment drywell space, stabilizes the direct-contact condensation in the pressure suppression pool and hampers onset of the chugging phenomenon, which induces most significant steam condensation load onto the pool boundary. This was found to be true for the tests with relatively small-break diameters, where the maximum steam mass fluxes in the vent pipe were lower than the upper threshold value for the onset of chugging. However, in the tests with the maximum vent steam mass fluxes moderately higher than the chugging upper threshold value, early depletion of the noncondensable gas tended to result in significant stabilization of steam condensation accompanied by an excursion of temperature of pool water surrounding the vent pipe outlets, which led to a delayed onset of chugging. Due to this combined influence of the noncondensable gas and nonuniform pool temperature, and due to dependence of magnitude of chugging load on the vent steam mass flux, the peak magnitude of the steam condensation load appearing in a blowdown can be very sensitive to the initial and break conditions.

  17. An analysis on the severe accident progression with operator recovery actions

    Energy Technology Data Exchange (ETDEWEB)

    Vo, T.H. [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejon 305-353 (Korea, Republic of); Korea University of Science and Technology (UST), 217 Gajeong-ro, Yuseong-gu, Daejeon 305-333 (Korea, Republic of); Song, J.H., E-mail: dosa@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejon 305-353 (Korea, Republic of); Korea University of Science and Technology (UST), 217 Gajeong-ro, Yuseong-gu, Daejeon 305-333 (Korea, Republic of); Kim, T.W.; Kim, D.H. [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejon 305-353 (Korea, Republic of)

    2014-12-15

    Highlights: • Severe accident progression for the station blackout and SBLOCA accident. • Analyses on APR1400 using MELCOR. • Operator recovery actions for decay heat removal and inventory make up. • Determine the time allowed for the operator to prevent reactor vessel failure. • Insight for the operator recovery actions for the severe accident management. - Abstract: Analyses on the severe accident progressions for the station blackout (SBO) accident and small break LOCA (SBLOCA) initiated severe accident were performed for APR1400 by using MELCOR computer code. Operator recovery actions for decay heat removal and inventory make up using a depressurization system and safety injection pump were simulated in parallel with a simulation of the severe accident progression. Sensitivity studies on the operator actions were performed to investigate the changes in the timing of the reactor vessel failure and to determine the time allowed for the operator to prevent reactor vessel failure. Sensitivity analyses on the effect of major modeling parameters were performed additionally to quantify the uncertainties in timing. It is found that the operator has about 2 h for the recovery actions after the indication of core damage by the signal of core exit thermocouple (CET) for the SBLOCA initiated severe accident, while the operator has to take immediate actions after the indication of core damage by CET for the SBO accident.

  18. Simulation of the aspersion system of the core at high pressure (HPCS) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de aspersion del nucleo alta presion (HPCS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas O, D.; Chavez M, C., E-mail: danmirnyi@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    A high-priority topic for the nuclear industry is the safety, consequently a nuclear power plant should have the emergency systems of cooling of the core (ECCS), designed exclusively to enter in operation in the event of an accident with coolant loss, including the design base accident. The objective of the aspersion system of the core at high pressure (HPCS) is to provide in an autonomous way the cooling to the core maintaining for if same the coolant inventory even when a small break is presented that does not allow the depressurization of the reactor and also avoiding excessive temperatures that affect the shielding of the fuel. The present work describes the development of the model and the simulation of the HPCS using the RELAP/SCDAP code. During the process simulation, for the setting in march of the system HPCS in an accident with coolant loss is necessary to implement the main components of the system taking into account what unites them, the main pump, the filled pump, the suction and injection valves, pipes and its water sources that can be condensed storage tanks and the suppression pool. The simulation of this system will complement the model with which counts the Analysis Laboratory in Nuclear Reactors Engineering of the UNAM regarding to the nuclear power plant of Laguna Verde which does not have a detailed simulation of the emergency cooling systems. (Author)

  19. Flow visualization study of inverted U-bend two-phase flow

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, M.; Kim, S.B.; Lee, R.

    1986-12-01

    A hot-leg U-bend experiment was performed. The experimental condition simulated the two-phase flow in a B and W primary loop during a small break loss of coolant accident or during some other abnormal transients. The loop design was based on the scaling criteria developed previously and the loop was operated either in a natural circulation mode or in a forced circulation mode using nitrogen gas and water. The two-phase flow regimes at the hot-leg were identified on the basis of visual observation. The phase separation at the top of the inverted U-bend was observed at low gas flow rate. The void fractions were measured using differential pressure transducers and compared with the prediction from the drift-flux model. The natural circulation flow interruption occurred in two different modes, namely, quasi-periodic and semi-permanent modes. This phenomenon is mainly dependent on the difference in the hydrostatic head in the riser and downcomer, and the flow regime at hot-leg. Besides this flow interruption phenomenon, dynamic flow instabilities of considerable amplitudes have been observed.

  20. An analytical and experimental investigation of natural circulation transients in a model pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Massoud, M

    1987-01-01

    Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS),'' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients.

  1. The simulation of thermohydraulic phenomena in a pressurized water reactor primary loop

    Energy Technology Data Exchange (ETDEWEB)

    Popp, M

    1987-01-01

    Several important fluid flow and heat transfer phenomena essential to nuclear power reactor safety were investigated. Scaling and modeling laws for pressurized water reactors are reviewed and a new scaling approach focusing on the overall loop behavior is presented. Scaling criteria for one- and two-phase natural circulation are developed, as well as a simplified model describing the first phase of a small break loss of coolant accident. Reactor vessel vent valve effects are included in the analysis of steady one-phase natural circulation flow. Two new dimensionless numbers, which uniquely describe one-phase flow in natural circulation loops, were deduced and are discussed. A scaled model of the primary loop of a typical Babcock and Wilcox reactor was designed, built, and tested. The particular prototype modeled was the TMI unit 2 reactor. The electrically heated, stainless steel model operates at a maximum pressure of 300 psig and has a maximum heat input of 188 kW. The model is about 4 times smaller in height than the prototype reactor, with a nominal volume scale of 1:500. Experiments were conducted establishing subcooled natural circulation in the model loop. Both steady flow and power transients were investigated.

  2. Source Term Analysis for the Nuclear Power Station Goesgen-Daeniken; Quelltermanalysen fuer das Kernkraftwerk Goesgen-Daeniken

    Energy Technology Data Exchange (ETDEWEB)

    Hosemann, J.P.; Megaritis, G.; Guentay, S.; Hirschmann, H.; Luebbesmeyer, D.; Lieber, K.; Jaeckel, B.; Birchley, J.; Duijvestijn, G

    2001-08-01

    Analyses are performed for three accident scenarios postulated to occur in the Goesgen Nuclear Power Plant, a 900 MWe Pressurised Water Reactor of Siemens design. The scenarios investigated comprise a Station Blackout and two separate cases of small break loss-of-coolant accident which lead, respectively, to high, intermediate and low pressure conditions in the reactor system. In each case the accident assumptions are highly pessimistic, so that the sequences span a large range of plant states and a damage phenomena. Thus the plant is evaluated for a diversity of potential safety challenges. A suite of analysis tools are used to examine the reactor coolant system response, the core heat-up, melting, fission product release from the reactor system, the transport and chemical behaviour of those fission products in the containment building, and the release of radioactivity (source term) to the environment. Comparison with reference values used by the licensing authority shows that the use of modern analysis tools and current knowledge can provide substantial reduction in the estimated source term. Of particular interest are insights gained from the analyses which indicate opportunities for operators to reduce or forestall the release. (author)

  3. Reactor safety issues resolved by the 2D/3D program

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated. This report was prepared in a coordination among US, Germany and Japan. US and Germany have published the report as NUREG/IA-0127 and GRS-101 respectively. (author).

  4. Simulation experiments for hot-leg U-bend two-phase flow phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, M.; Hsu, J.T.; Tucholke, D.; Lambert, G.; Kataoka, I.

    1986-01-01

    In order to study the two-phase natural circulation and flow termination during a small break loss of coolant accident in LWR, simulation experiments have been performed. Based on the two-phase flow scaling criteria developed under this program, an adiabatic hot leg U-bend simulation loop using nitrogen gas and water and a Freon 113 boiling and condensation loop were built. The nitrogen-water system has been used to isolate key hydrodynamic phenomena from heat transfer problems, whereas the Freon loop has been used to study the effect of phase changes and fluid properties. Various tests were carried out to establish the basic mechanism of the flow termination and reestablishment as well as to obtain essential information on scale effects of parameters such as the loop frictional resistance, thermal center, U-bend curvature and inlet geometry. In addition to the above experimental study, a preliminary modeling study has been carried out for two-phase flow in a large vertical pipe at relatively low gas fluxes typical of natural circulation conditions.

  5. Model Description of TASS/SMR Code

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Y. D.; Yang, S. H.; Kim, S. H.; Lee, S. W.; Kim, H. K.; Yoon, H. Y.; Lee, G. H.; Bae, K. H.; Chung, Y. J

    2005-12-15

    TASS/SMR(Transient And Setpoint Simulation/System-integrated Modular Reactor) code has been developed for the safety analysis of the SMART-P reactor. TASS/SMR code can be applied for the analysis of design base accidents including small break loss of coolant accident of the SMART research reactor. TASS/SMR code models the primary and secondary system using a node and flow path. A node represents the control volume which defines the fluid mass and energy. Flow path connects the nodes to define the momentum of the fluid. The mass and energy conservation equations are applied to the node and the momentum conservation equation applied to the flow path. In TASS/SMR, the governing equations are applied for both the primary and the secondary coolant system and are solved simultaneously. The governing equations of TASS/SMR are based on the drift-flux model so that the accidents or transients accompanying with two-phase flow can be analyzed. Also, the SMART-P reactor specific thermal-hydraulic models are incorporated, such as non-condensable gas model, helical steam generator heat transfer model, and passive residual heat removal system (PRHRS) heat transfer model. This technical report describes the governing equations, solution method, thermal hydraulics, reactor core, control system models used in TASS/SMR code. Also, the description for the steady state simulation, the minimum CHFR and hottest fuel temperature calculation methods are described in this report.

  6. Preliminary Test of a small heat pipe for hybrid control rod in-core passive decay heat removal system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In Guk; Ban, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-05-15

    This paper introduces 'Hybrid control rod' combining its original function and heat removal ability. The high temperature operation and high resistance of radiation should be considered to adopt the hybrid heat pipe at the in-core condition. Other design consideration is to make extra inlet parts because it has a high risk of inlet boundary failure. It means that the introduction of heat pipe system is difficult to present nuclear power plants. The other concepts are presented to out-core cooling design but it has low performance compared with in-core heat removal system. Hybrid heat pipe for in-core heat removal system suggests the solution of these problems. Ultimate objective of this research is to develop the passive emergency decay heat removal system using hybrid heat pipes targeting design bases accidents such as station black-out (SBO) and small break loss of coolant accident (SBLOCA). The purpose of this work is to confirm the performance and heat transfer behavior of hybrid heat pipe. The hybrid heat pipe has special condition for operation. Therefore, it is hard to analyze their behavior in core. Table I shows the characteristics of hybrid heat pipe and consideration for manufacturing the heat pipe.

  7. Major Achievements and Prospect of the ATLAS Integral Effect Tests

    Directory of Open Access Journals (Sweden)

    Ki-Yong Choi

    2012-01-01

    Full Text Available A large-scale thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation, has been operated by KAERI. The reference plant of ATLAS is the APR1400 (Advanced Power Reactor, 1400 MWe. Since 2007, an extensive series of experimental works were successfully carried out, including large break loss of coolant accident tests, small break loss of coolant accident tests at various break locations, steam generator tube rupture tests, feed line break tests, and steam line break tests. These tests contributed toward an understanding of the unique thermal-hydraulic behavior, resolving the safety-related concerns and providing validation data for evaluation of the safety analysis codes and methodology for the advanced pressurized water reactor, APR1400. Major discoveries and lessons found in the past integral effect tests are summarized in this paper. As the demand for integral effect tests is on the rise due to the active national nuclear R&D program in Korea, the future prospects of the application of the ATLAS facility are also discussed.

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

    Directory of Open Access Journals (Sweden)

    Virpi Kouhia

    2012-01-01

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

  9. SCDAP/RELAP5/MOD2 code manual

    Energy Technology Data Exchange (ETDEWEB)

    Allison, C.M.; Johnson, E.C. (eds.); Berna, G.A.; Cheng, T.C.; Hagrman, D.L.; Johnsen, G.W.; Kiser, D.M.; Miller, C.S.; Ransom, V.H.; Riemke, R.A.; Shieh, A.S.; Siefken, L.J.; Trapp, J.A.; Wagner, R.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1989-09-01

    The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, and the fission products and aerosols in the system during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This document, Volume III, contains detailed instructions for code application and input data preparation. In addition, Volume III contains user guidelines that have evolved over the past several years from application of the RELAP5 and SCDAP codes at the Idaho National Engineering Laboratory, at other national laboratories, and by users throughout the world. 2 refs., 32 figs., 9 tabs.

  10. Establishment of safety analysis system for emergency operating procedures of Kori 3 and 4 and YGN 1 and 2 and reference analyses

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Kim, Hee Cheol; Ha, Kwi Seok; Chung, Bub Dong; Jeong, Jae Jun

    1999-02-01

    This report describes the establishment of safety analysis system for emergency operating procedures(EOPs) of Kori 3 and 4 and YGN 1 and 2 and the results of reference analyses. MARS1.3.1 code has been selected as a realistic system analysis tool of the safety analysis system, and the reactor system has been modeled using a fine nodding scheme in order to capture the major thermal-hydraulic phenomena that might occur during the transients. Full power steady state operating conditions are generated based on the plant operation data. Then, the reference analyses have been carried out for the accidents that can represent the typical EOPtransients, that is, a small-break loss-of-coolant-accident,a main steam line break accident and a steam generator tube rupture accident from the full power operation. For the realistic simulation of plant transient responses, reactor control and protection systems and safety systems are modeled based on their realistic performances. Also, the operator actions are modeled based on the current EOP actions. Through the reference analyses, the soundness of the established safety analysis system and the system modeling has been verified and the effectiveness of the current EOP has been partly justified. In conclusion, the safety analysis system established through this study can be used for the generation of technical background in the development and improvement of EOP actions and in the operator training. (Author). 11 refs., 6 tabs., 48 figs.

  11. Gest-sip1 experiments and post-test calculations with the relap5 code

    Energy Technology Data Exchange (ETDEWEB)

    Achilli, A.; Cattadori, G.; Ferri, R.; Gandolfi, S. [SIET S.p.A., Piacenza (Italy); Bianchi, F.; Meloni, P. [ENEA - Centro Ricerche ' ' Ezio Clementel' ' , Bologna (Italy)

    2001-07-01

    The SIP-1 apparatus (Sistema di Iniezione Passiva) was conceived, designed, numerically simulated and tested by the SIET company as an innovative depressurization and make-up device for the New Generation LWRs. In particular it is suitable to cope with those accidents where pressure in the circuit must be dumped to allow low pressure injection systems to intervene. The main peculiarity of SIP-1 is the capability of de-pressurizing a system by cold water injection, rather than by discharging mass to the outlet, as in the common depressurization systems. ENEA sponsored all the research activity, starting from the SIP-1 design, its numerical simulation with the Relap5 code, the realisation of an experimental facility up to the test execution and post-test calculations. An experimental campaign on the GEST-SIP1 facility was performed in July 2000. The facility is mainly constituted by a U-tube Steam Generator which a proper model of SIP-1 apparatus is connected to. A series of Small Break LOCAs was simulated by varying the break size and different steady conditions were investigated to verify the stability of SIP-1, the lack of unexpected interventions and the actuation modalities. This paper deals with the description of the GEST-SIP1 experimental facility, the SIP-1 operating principles, the most meaningful results of the tests and the capability of the Relap5 code in reproducing phenomena and events. (author)

  12. TRAC-PF1/MOD1 post-test calculations of the OECD (Organisation for Economic Co-operation and Development) LOFT experiment LP-SB-2

    Energy Technology Data Exchange (ETDEWEB)

    Pelayo, F. (UKAEA Atomic Energy Establishment, Winfrith (UK))

    1990-12-01

    An analysis of the OECD-LOFT-LP-SB-2 experiment making use of TRAC-PF1/MOD1 is described in the report. LP-SB2 experiment studies the effect of a delayed pump trip in a small break LOCA scenario with a 3 inches equivalent diameter break in the hot leg of a commercial PWR operating at full power. The experiment was performed on 14 July 1983 in the LOFT facility at the Idaho National Engineering Laboratory under the auspices of the Organization for Economic Co-operation and Development (OECD). This analysis presents an evaluation of the code capability in reproducing the complex phenomena which determined the LP-SB-2 transient evolution. the analysis comprises the results obtained from two different runs. The first run is described in detail analysing the main variables over two time spans: short and longer term. Several conclusions are drawn and then a second run testing some of these conclusions is shown. All of the calculations were preformed at the United Kingdom Atomic Energy Establishment at Winfrith under the auspices of an agreement between the UKAEA (United Kingdom Atomic Energy Authority) and the Consejo de Seguridad Nuclear Espanol (CSN). 16 refs., 64 figs., 6 tabs.

  13. Assessment of RELAP5/MOD3 with the LOFT L9-1/L3-3 experiment simulating an anticipated transient with multiple failures

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Y.S.; Seul, K.W.; Kim, H.J. [Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

    1994-02-01

    The RELAP5/MOD3 5m5 code is assessed using the L9-1/L3-3 test carried out in the LOFT facility, a 1/60-scaled experimental reactor, simulating a loss of feedwater accident with multiple failures and the sequentially-induced small break loss-of-coolant accident. The code predictability is evaluated for the four separated sub-periods with respect to the system response; initial heatup phase, spray and power operated relief valve (PORV) cycling phase, blowdown phase and recovery phase. Based on the comparisons of the results from the calculation with the experiment data, it is shown that the overall thermal-hydraulic behavior important to the scenario such as a heat removal between the primary side and the secondary side and a system depressurization can be well-predicted and that the code could be applied to the full-scale nuclear power plant for an anticipated transient with multiple failures within a reasonable accuracy. The minor discrepancies between the prediction and the experiment are identified in reactor scram time, post-scram behavior in the initial heatup phase, excessive heatup rate in the cycling phase, insufficient energy convected out the PORV under the hot leg stratified condition in the saturated blowdown phase and void distribution in secondary side in the recovery phase. This may come from the code uncertainties in predicting the spray mass flow rate, the associated condensation in pressurizer and junction fluid density under stratified condition.

  14. SCDAP/RELAP5/MOD2 code manual

    Energy Technology Data Exchange (ETDEWEB)

    Allison, C.M.; Johnson, E.C. (eds.); Berna, G.A.; Cheng, T.C.; Hagrman, D.L.; Johnsen, G.W.; Kiser, D.M.; Miller, C.S.; Ransom, V.H.; Riemke, R.A.; Shieh, A.S.; Siefken, L.J.; Trapp, J.A.; Wagner, R.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1989-09-01

    The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, and the fission products and aerosols in the system during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. The modeling theory and associated numerical schemes are documented in Volumes I and in this document, Volume II, to acquaint the user with the modeling base and thus aid in effective use of the code. 135 refs., 48 figs., 8 tabs.

  15. Consistent Posttest Calculations for LOCA Scenarios in LOBI Integral Facility

    Directory of Open Access Journals (Sweden)

    F. Reventós

    2012-01-01

    Full Text Available Integral test facilities (ITFs are one of the main tools for the validation of best estimate thermalhydraulic system codes. The experimental data are also of great value when compared to the experiment-scaled conditions in a full NPP. The LOBI was a single plus a triple-loop (simulated by one loop test facility electrically heated to simulate a 1300 MWe PWR. The scaling factor was 712 for the core power, volume, and mass flow. Primary and secondary sides contained all main active elements. Tests were performed for the characterization of phenomenologies relevant to large and small break LOCAs and special transients in PWRs. The paper presents the results of three posttest calculations of LOBI experiments. The selected experiments are BL-30, BL-44, and A1-84. They are LOCA scenarios of different break sizes and with different availability of safety injection components. The goal of the analysis is to improve the knowledge of the phenomena occurred in the facility in order to use it in further studies related to qualifying nodalizations of actual plants or to establish accuracy data bases for uncertainty methodologies. An example of procedure of implementing changes in a common nodalization valid for simulating tests occurred in a specific ITF is presented along with its confirmation based on posttests results.

  16. The role of the uncertainty in code development

    Energy Technology Data Exchange (ETDEWEB)

    Barre, F. [CEA-Grenoble (France)

    1997-07-01

    From a general point of view, all the results of a calculation should be given with their uncertainty. It is of most importance in nuclear safety where sizing of the safety systems, therefore protection of the population and the environment essentially depends on the calculation results. Until these last years, the safety analysis was performed with conservative tools. Two types of critics can be made. Firstly, conservative margins can be too large and it may be possible to reduce the cost of the plant or its operation with a best estimate approach. Secondly, some of the conservative hypotheses may not really conservative in the full range of physical events which can occur during an accident. Simpson gives an interesting example: in some cases, the majoration of the residual power during a small break LOCA can lead to an overprediction of the swell level and thus of an overprediction of the core cooling, which is opposite to a conservative prediction. A last question is: does the accumulation of conservative hypotheses for a problem always give a conservative result? The two phase flow physics, mainly dealing with situation of mechanical and thermal non-equilibrium, is too much complicated to answer these questions with a simple engineer judgement. The objective of this paper is to make a review of the quantification of the uncertainties which can be made during code development and validation.

  17. 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.

  18. MELCOR simulation of postulated severe accidents in OPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seongn Yeon; Kim Sung Joong [Hanyang Univ., Seoul (Korea, Republic of); Kim, Hwan Yeol; Park, Jong Hwa [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    Since the Fukushima accident in 2011, severe accidents of a nuclear power plant have been a target of big debate whether the defense in depth philosophy applied to current nuclear system is still vigorous enough to ensure the protection of the operators and the public. Thus an accurate prediction of severe accident has become a critical task for the nuclear engineers with reliable employment of Probabilistic Risk Analysis (PRA). According to a recent PRA result, Small Break Loss Of Coolant Accident (SBLOCA) without safety injection and Station Black Out (SBO) show high probability of proceeding to severe accidents. Thus, these accident scenarios need to be evaluated properly with reliable prediction tools. Song and Ahn analyzed SBO sequences in KSNP using MELCOR 1.8.5. Park and Song examined SBLOCA scenarios based on the PSA of KNSP using MAAP 4.06. Their studies utilized severe accident database. In continuation of the further analysis, several scenarios of postulated SBO and SBLOCA in OPR1000 are investigated using the severe accident database and MELCOR 1.8.6.

  19. A Randomized Controlled Trial of Team-Based Learning Versus Lectures with Break-Out Groups on Knowledge Retention.

    Science.gov (United States)

    Thrall, Grace C; Coverdale, John H; Benjamin, Sophiya; Wiggins, Anna; Lane, Christianne Joy; Pato, Michele T

    2016-10-01

    This goal of this study was to evaluate the efficacy of team-based learning (TBL) on knowledge retention compared to traditional lectures with small break-out group discussion (teaching as usual (TAU)) using a randomized controlled trial. This randomized controlled trial was conducted during a daylong conference for psychiatric educators on attention-deficit hyperactivity disorder and the research literacy topic of efficacy versus effectiveness trials. Learners (n = 115) were randomized with concealed allocation to either TBL or TAU. Knowledge was measured prior to the intervention, immediately afterward, and 2 months later via multiple-choice tests. Participants were necessarily unblinded. Data enterers, data analysts, and investigators were blinded to group assignment in data analysis. Per-protocol analyses of test scores were performed using change in knowledge from baseline. The primary endpoint was test scores at 2 months. At baseline, there were no statistically significant differences between groups in pre-test knowledge. At immediate post-test, both TBL and TAU groups showed improved knowledge scores compared with their baseline scores. The TBL group performed better statistically on the immediate post-test than the TAU group (Cohen's d = 0.73; p < 0.001), although the differences in knowledge scores were not educationally meaningful, averaging just one additional test question correct (out of 15). On the 2-month remote post-test, there were no group differences in knowledge retention among the 42 % of participants who returned the 2-month test. Both TBL and TAU learners acquired new knowledge at the end of the intervention and retained knowledge over 2 months. At the end of the intervention day and after 2 months, knowledge test scores were not meaningfully different between TBL and TAU completers. In conclusion, this study failed to demonstrate the superiority of TBL over TAU on the primary outcome of knowledge retention at 2 months post-intervention.

  20. Podolsky electromagnetism and a modification in Stefan-Boltzmann law

    Energy Technology Data Exchange (ETDEWEB)

    Bonin, Carlos Alberto; Bufalo, Rodrigo Santos; Escobar, Bruto Max Pimentel; Zambrano, German Enrique Ramos [Instituto de Fisica Teorica (IFT/UNESP), Sao Paulo, SP (Brazil)

    2009-07-01

    Full text. As it is well-known, gauge fields that emerge from the gauge principle are massless vector fields. Considering the photon as a Proca particle, experience sets an upper limit on its mass. This limit is m{sub Proca} < 6X10{sup -17}eV (PDG 2006). However, a mass term, regardless how small, breaks the gauge symmetry. Nevertheless, there exists a theory in which is possible to introduce a mass term preserving all symmetries of Maxwell electromagnetism, including the gauge one: such theory is known as Podolsky Electromagnetism. Podolsky theory is a second- order-derivative theory and has some remarkable properties, despite those already mentioned: the theory has two sectors, a massive one and massless one, it depends on a free parameter (which happens to be the mass of the massive sector) that, like all other elementary particles's masses of the Standard Model, must be fixed through experiences, and the fact that the electrostatic potential is finite everywhere, including over a punctual charge. Just like Maxwell electromagnetism, Podolsky's is a constrained theory and, since it is of second order in the derivatives, it consists in a much richer theoretical structure. Therefore, from both, theoretical and experimental points of view, Podolsky electromagnetism is a very attractive theory. In this work we study a gas of Podolsky photons at finite temperature through path integration. We show that the massless sector leads to the famous Planck's law for black-body radiation and, therefore, to the Stefan-Boltzmann law. We also show that the massive sector of the Podolsky theory induces a modification in both these laws. It is possible to set limits on the Podolsky parameter through comparison of our results with data from cosmic microwave background radiation. (author)

  1. Analysis of Loss-of-Coolant Accidents in the NBSR

    Energy Technology Data Exchange (ETDEWEB)

    Baek J. S.; Cheng L.; Diamond, D.

    2014-05-23

    This report documents calculations of the fuel cladding temperature during loss-of-coolant accidents in the NBSR. The probability of a pipe failure is small and procedures exist to minimize the loss of water and assure emergency cooling water flows into the reactor core during such an event. Analysis in the past has shown that the emergency cooling water would provide adequate cooling if the water filled the flow channels within the fuel elements. The present analysis is to determine if there is adequate cooling if the water drains from the flow channels. Based on photographs of how the emergency water flows into the fuel elements from the distribution pan, it can be assumed that this water does not distribute uniformly across the flow channels but rather results in a liquid film flowing downward on the inside of one of the side plates in each fuel element and only wets the edges of the fuel plates. An analysis of guillotine breaks shows the cladding temperature remains below the blister temperature in fuel plates in the upper section of the fuel element. In the lower section, the fuel plates are also cooled by water outside the element that is present due to the hold-up pan and temperatures are lower than in the upper section. For small breaks, the simulation results show that the fuel elements are always cooled on the outside even in the upper section and the cladding temperature cannot be higher than the blister temperature. The above results are predicated on assumptions that are examined in the study to see their influence on fuel temperature.

  2. Reactor vessel lower head integrity

    Energy Technology Data Exchange (ETDEWEB)

    Rubin, A.M.

    1997-02-01

    On March 28, 1979, the Three Mile Island Unit 2 (TMI-2) nuclear power plant underwent a prolonged small break loss-of-coolant accident that resulted in severe damage to the reactor core. Post-accident examinations of the TMI-2 reactor core and lower plenum found that approximately 19,000 kg (19 metric tons) of molten material had relocated onto the lower head of the reactor vessel. Results of the OECD TMI-2 Vessel Investigation Project concluded that a localized hot spot of approximately 1 meter diameter had existed on the lower head. The maximum temperature on the inner surface of the reactor pressure vessel (RPV) in this region reached 1100{degrees}C and remained at that temperature for approximately 30 minutes before cooling occurred. Even under the combined loads of high temperature and high primary system pressure, the TMI-2 RPV did not fail. (i.e. The pressure varied from about 8.5 to 15 MPa during the four-hour period following the relocation of melt to the lower plenum.) Analyses of RPV failure under these conditions, using state-of-the-art computer codes, predicted that the RPV should have failed via local or global creep rupture. However, the vessel did not fail; and it has been hypothesized that rapid cooling of the debris and the vessel wall by water that was present in the lower plenum played an important role in maintaining RPV integrity during the accident. Although the exact mechanism(s) of how such cooling occurs is not known, it has been speculated that cooling in a small gap between the RPV wall and the crust, and/or in cracks within the debris itself, could result in sufficient cooling to maintain RPV integrity. Experimental data are needed to provide the basis to better understand these phenomena and improve models of RPV failure in severe accident codes.

  3. Design of Reactor Coolant Pump Seal Online Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Ah, Sang Ha; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of); Lee, Song Kyu [Korea Power Engineering Co., Yongin (Korea, Republic of)

    2008-05-15

    As a part of a Department of Korea Power Engineering Co., (KOPEC) Project, Statistical Quality Control techniques have been applied to many aspects of industrial engineering. An application to nuclear power plant maintenance and control is also presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) and the fouling resistance of heat exchanger. This research uses Shewart X-bar, R charts, Cumulative Sum charts (CUSUM), and Sequential Probability Ratio Test (SPRT) to analyze the process for the state of statistical control. And the Control Chart Analyzer (CCA) has been made to support these analyses that can make a decision of error in process. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with enough time to respond to possible emergency situations and thus improve plant safety and reliability. RCP circulates reactor coolant to transfer heat from the reactor to the steam generators. RCP seals are in the pressure part of reactor coolant system, so if it breaks, it can cause small break LOCA. And they are running on high pressure, and high temperature, so they can be easily broken. Since the reactor coolant pumps operate within the containment building, physical access to the pumps occurs only during refueling outages. Engineers depend on process variables transmitted to the control room and through the station's data historian to assess the pumps' condition during normal operation.

  4. Pre-Study of Off-site Consequence Analysis in Level 3 PSA of Wolsong Unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won-Jik; Yang, Ho-Chang; Choi, Seong-Soo [ACT, Daejeon (Korea, Republic of)

    2015-10-15

    In order to perform level 3 PSA, MACCS II (MELCOR Accident Consequence Code System 2) is needed. MACCS II is used in PSA for plants in order to evaluate population dose that is the effects on health and environment caused by released radioisotopes after an accident. In this study, Steam Generator Tube Rupture (SGTR) event in CANDU-6 plants is evaluated population dose that is the effects on health and environment caused by released radioisotopes after an accident. In this study, Steam Generator Tube Rupture (SGTR) event has been evaluated by using Level 1 PSA result and Level 2 PSA result(ISSAC) and MACCS II. As a result, We are obtained the following conclusion. - Early maximum early fatalities is 5.35E+02 equal to latent maximum early fatalities.(99.5%) - Early and latent maximum cancer fatalities are 2.33E+03 and 1.11E+04, respectively. (99.5%) - Early and latent maximum population doses are 1.25 and 5.00 person-rem/yr, respectively. (99.5%) Other study has shown that MACCS II was performed evaluation for Wolsong NPP. Small Break Loss of Coolant Accident(SBLOCA) event is selected by other study. The results of early and cancer fatalities applied similar assumption were 3.02E+00 and 1.89E+03, respectively. This study's results are higher than other study's result. Because, basis input data is different each studies, and event frequency are different (This study : 2.10E-07/ Other study : 4.93E-09)

  5. Preliminary safety analysis methodology for the SMART

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Kyoo Hwan; Chung, Y. J.; Kim, H. C.; Sim, S. K.; Lee, W. J.; Chung, B. D.; Song, J. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    This technical report was prepared for a preliminary safety analysis methodology of the 330MWt SMART (System-integrated Modular Advanced ReacTor) which has been developed by Korea Atomic Energy Research Institute (KAERI) and funded by the Ministry of Science and Technology (MOST) since July 1996. This preliminary safety analysis methodology has been used to identify an envelope for the safety of the SMART conceptual design. As the SMART design evolves, further validated final safety analysis methodology will be developed. Current licensing safety analysis methodology of the Westinghouse and KSNPP PWRs operating and under development in Korea as well as the Russian licensing safety analysis methodology for the integral reactors have been reviewed and compared to develop the preliminary SMART safety analysis methodology. SMART design characteristics and safety systems have been reviewed against licensing practices of the PWRs operating or KNGR (Korean Next Generation Reactor) under construction in Korea. Detailed safety analysis methodology has been developed for the potential SMART limiting events of main steam line break, main feedwater pipe break, loss of reactor coolant flow, CEA withdrawal, primary to secondary pipe break and the small break loss of coolant accident. SMART preliminary safety analysis methodology will be further developed and validated in parallel with the safety analysis codes as the SMART design further evolves. Validated safety analysis methodology will be submitted to MOST as a Topical Report for a review of the SMART licensing safety analysis methodology. Thus, it is recommended for the nuclear regulatory authority to establish regulatory guides and criteria for the integral reactor. 22 refs., 18 figs., 16 tabs. (Author)

  6. Multi-dimensional boron transport modeling in subchannel approach: Part II. Validation of CTF boron tracking model and adding boron precipitation model

    Energy Technology Data Exchange (ETDEWEB)

    Ozdemir, Ozkan Emre, E-mail: ozdemir@psu.edu; Avramova, Maria N., E-mail: mna109@psu.edu

    2014-10-15

    Highlights: • Validation of implemented multi-dimensional subchannel boron transport model. • Extension of boron transport model to entrained droplets. • Implementation of boron precipitation model. • Testing of the boron precipitation model under transient condition. - Abstract: The risk of small-break loss of coolant accident (SB-LOCA) and other reactivity initiated transients caused by boron dilution in the light water reactors (LWRs), and the complications of tracking the soluble boron concentration experimentally inside the primary coolant have stimulated the interest in computational studies for accurate boron tracking simulations in nuclear reactors. In Part I of this study, the development and implementation of a multi-dimensional boron transport model with modified Godunov scheme based on a subchannel approach within the COBRA-TF (CTF) thermal-hydraulic code was presented. The modified Godunov scheme approach with a physical diffusion term was determined to provide the most accurate and precise solution. Current paper extends these conclusions and presents the model validation studies against experimental data from the Rossendorf coolant mixing model (ROCOM) test facility. In addition, the importance of the two-phase flow characteristics in modeling boron transient are emphasized, especially during long-term cooling period after the loss of coolant accident (LOCA) condition in pressurized water reactors (PWRs). The CTF capabilities of boron transport modeling are further improved based on the three-field representation of the two-phase flow utilized in the code. The boron transport within entrained droplets is modeled, and a model for predicting the boron precipitation under transient conditions is developed and tested. It is aimed to extend the applicability of CTF to reactor transient simulations, and particularly to a large-break loss of coolant accident (LB-LOCA) analysis.

  7. Uncertainty Analysis of RELAP5-3D

    Energy Technology Data Exchange (ETDEWEB)

    Alexandra E Gertman; Dr. George L Mesina

    2012-07-01

    As world-wide energy consumption continues to increase, so does the demand for the use of alternative energy sources, such as Nuclear Energy. Nuclear Power Plants currently supply over 370 gigawatts of electricity, and more than 60 new nuclear reactors have been commissioned by 15 different countries. The primary concern for Nuclear Power Plant operation and lisencing has been safety. The safety of the operation of Nuclear Power Plants is no simple matter- it involves the training of operators, design of the reactor, as well as equipment and design upgrades throughout the lifetime of the reactor, etc. To safely design, operate, and understand nuclear power plants, industry and government alike have relied upon the use of best-estimate simulation codes, which allow for an accurate model of any given plant to be created with well-defined margins of safety. The most widely used of these best-estimate simulation codes in the Nuclear Power industry is RELAP5-3D. Our project focused on improving the modeling capabilities of RELAP5-3D by developing uncertainty estimates for its calculations. This work involved analyzing high, medium, and low ranked phenomena from an INL PIRT on a small break Loss-Of-Coolant Accident as wall as an analysis of a large break Loss-Of- Coolant Accident. Statistical analyses were performed using correlation coefficients. To perform the studies, computer programs were written that modify a template RELAP5 input deck to produce one deck for each combination of key input parameters. Python scripting enabled the running of the generated input files with RELAP5-3D on INL’s massively parallel cluster system. Data from the studies was collected and analyzed with SAS. A summary of the results of our studies are presented.

  8. Pilot study of dynamic Bayesian networks approach for fault diagnostics and accident progression prediction in HTR-PM

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Yunfei; Tong, Jiejuan; Zhang, Liguo, E-mail: lgzhang@tsinghua.edu.cn; Zhang, Qin

    2015-09-15

    Highlights: • Dynamic Bayesian network is used to diagnose and predict accident progress in HTR-PM. • Dynamic Bayesian network model of HTR-PM is built based on detailed system analysis. • LOCA Simulations validate the above model even if part monitors are lost or false. - Abstract: The first high-temperature-reactor pebble-bed demonstration module (HTR-PM) is under construction currently in China. At the same time, development of a system that is used to support nuclear emergency response is in progress. The supporting system is expected to complete two tasks. The first one is diagnostics of the fault in the reactor based on abnormal sensor measurements obtained. The second one is prognostic of the accident progression based on sensor measurements obtained and operator actions. Both tasks will provide valuable guidance for emergency staff to take appropriate protective actions. Traditional method for the two tasks relies heavily on expert judgment, and has been proven to be inappropriate in some cases, such as Three Mile Island accident. To better perform the two tasks, dynamic Bayesian networks (DBN) is introduced in this paper and a pilot study based on the approach is carried out. DBN is advantageous in representing complex dynamic systems and taking full consideration of evidences obtained to perform diagnostics and prognostics. Pearl's loopy belief propagation (LBP) algorithm is recommended for diagnostics and prognostics in DBN. The DBN model of HTR-PM is created based on detailed system analysis and accident progression analysis. A small break loss of coolant accident (SBLOCA) is selected to illustrate the application of the DBN model of HTR-PM in fault diagnostics (FD) and accident progression prognostics (APP). Several advantages of DBN approach compared with other techniques are discussed. The pilot study lays the foundation for developing the nuclear emergency response supporting system (NERSS) for HTR-PM.

  9. Numerical simulation of radioisotope's dependency on containment performance for large dry PWR containment under severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Mehboob, Khurram, E-mail: khurramhrbeu@gmail.com [College of Nuclear Science and Technology, Harbin Engineering University, 145-31 Nantong Street, Nangang District, Harbin, Heilongjiang 150001 (China); Xinrong, Cao [College of Nuclear Science and Technology, Harbin Engineering University, 145-31 Nantong Street, Nangang District, Harbin, Heilongjiang 150001 (China); Ahmed, Raheel [College of Automation, Harbin Engineering University, 145-31 Nantong Street, Nangang District, Harbin, Heilongjiang 150001 (China); Ali, Majid [College of Nuclear Science and Technology, Harbin Engineering University, 145-31 Nantong Street, Nangang District, Harbin, Heilongjiang 150001 (China)

    2013-09-15

    Highlights: • Calculation and comparison of activity of BURN-UP code with ORIGEN2 code. • Development of SASTC computer code. • Radioisotopes dependency on containment ESFs. • Mitigation in atmospheric release with ESFs operation. • Variation in radioisotopes source term with spray flow and pH value. -- Abstract: During the core melt accidents large amount of fission products can be released into the containment building. These fission products escape into the environment to contribute in accident source term. The mitigation in environmental release is demanded for such radiological consequences. Thus, countermeasures to source term, mitigations of release of radioactivity have been studied for 1000 MWe PWR reactor. The procedure of study is divided into five steps: (1) calculation and verification of core inventory, evaluated by BURN-UP code, (2) containment modeling based on radioactivity removal factors, (3) selection of potential accidents initiates the severe accident, (4) calculation of release of radioactivity, (5) study the dependency of release of radioactivity on containment engineering safety features (ESFs) inducing mitigation. Loss of coolant accident (LOCA), small break LOCA and flow blockage accidents (FBA) are selected as initiating accidents. The mitigation effect of ESFs on source term has been studied against ESFs performance. Parametric study of release of radioactivity has been carried out by modeling and simulating the containment parameters in MATLAB, which takes BURN-UP outcomes as input along with the probabilistic data. The dependency of iodine and aerosol source term on boric and caustic acid spray has been determined. The variation in source term mitigation with the variation of containment spray flow rate and pH values have been studied. The variation in containment retention factor (CRF) has also been studied with the ESF performance. A rapid decrease in source term is observed with the increase in pH value.

  10. A concept of JAERI passive safety light water reactor system (JPSR)

    Energy Technology Data Exchange (ETDEWEB)

    Murao, Y.; Araya, F.; Iwamura, T. [Japan Atomic Energy Research Institute, Tokai-mura (Japan)

    1995-09-01

    The Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor system concept, JPSR, which was developed for reducing manpower in operation and maintenance and influence of human errors on reactor safety. In the concept the system was extremely simplified. The inherent matching nature of core generation and heat removal rate within a small volume change of the primary coolant is introduced by eliminating chemical shim and adopting in-vessel control rod drive mechanism units, a low power density core and once-through steam generators. In order to simplify the system, a large pressurizer, canned pumps, passive engineered-safety-features-system (residual heat removal system and coolant injection system) are adopted and the total system can be significantly simplified. The residual heat removal system is completely passively actuated in non-LOCAs and is also used for depressurization of the primary coolant system to actuate accumulators in small break LOCAs and reactor shutdown cooling system in normal operation. All of systems for nuclear steam supply system are built in the containment except for the air coolers as a the final heat sink of the passive residual heat removal system. Accordingly the reliability of the safety system and the normal operation system is improved, since most of residual heat removal system is always working and a heat sink for normal operation system is {open_quotes}safety class{close_quotes}. In the passive coolant injection system, depressurization of the primary cooling system by residual heat removal system initiates injection from accumulators designed for the MS-600 in medium pressure and initiates injection from the gravity driven coolant injection pool at low pressure. Analysis with RETRAN-02/MOD3 code demonstrated the capability of passive load-following, self-power-controllability, cooling and depressurization.

  11. Modelling of WWER-1000 steam generators by REALP5/MOD3.2 code

    Energy Technology Data Exchange (ETDEWEB)

    D`Auria, F.; Galassi, G.M. [Univ. of Pisa (Italy); Frogheri, M. [Univ. of Genova (Italy)

    1997-12-31

    The presentation summarises the results of best estimate calculations carried out with reference to the WWER-1000 Nuclear Power Plant, utilizing a qualified nodalization set-up for the Relap5/Mod3.2 code. The nodalization development has been based on the data of the Kozloduy Bulgarian Plant. The geometry of the steam generator imposed drastic changes in noding philosophy with respect to what is suitable for the U-tubes steam generators. For the secondary side a symmetry axis was chosen to separate (in the nodalization) the hot and the cold sides of the tubes. In this way the secondary side of the steam generators was divided into three zones: (a) the hot zone including the hot collector and the hot l/2 parts of the tubes; (b) the cold zone including the cold collector and the cold 1/2 parts of the tubes; (c) the downcomer region, where down flow is assumed. As a consequence of above in the primary side more nodes are placed on the hot side of the tubes. Steady state and transient qualification has been achieved, considering the criteria proposed at the University of Pisa, utilizing plant transient data from the Kozloduy and the Ukrainian Zaporosche Plants. The results of the application of the qualified WWER-1000 Relap5/Mod3.2 nodalization to various transients including large break LOCA, small break LOCA and steam generator tube rupture, together with a sensitivity analysis on the steam generators, are reported in the presentation. Emphasis is given to the prediction of the steam generators performances. 23 refs.

  12. Validation and application of the system code ATHLET-CD for BWR severe accident analyses

    Energy Technology Data Exchange (ETDEWEB)

    Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Imke, Uwe; Sanchez, Victor

    2016-10-15

    Highlights: • We present the application of the system code ATHLET-CD code for BWR safety analyses. • Validation of core in-vessel models is performed based on KIT CORA experiments. • A SB-LOCA scenario is simulated on a generic German BWR plant up to vessel failure. • Different core reflooding possibilities are investigated to mitigate the accident consequences. • ATHLET-CD modelling features reflect the current state of the art of severe accident codes. - Abstract: This paper is aimed at the validation and application of the system code ATHLET-CD for the simulation of severe accident phenomena in Boiling Water Reactors (BWR). The corresponding models for core degradation behaviour e.g., oxidation, melting and relocation of core structural components are validated against experimental data available from the CORA-16 and -17 bundle tests. Model weaknesses are discussed along with needs for further code improvements. With the validated ATHLET-CD code, calculations are performed to assess the code capabilities for the prediction of in-vessel late phase core behaviour and reflooding of damaged fuel rods. For this purpose, a small break LOCA scenario for a generic German BWR with postulated multiple failures of the safety systems was selected. In the analysis, accident management measures represented by cold water injection into the damaged reactor core are addressed to investigate the efficacy in avoiding or delaying the failure of the reactor pressure vessel. Results show that ATHLET-CD is applicable to the description of BWR plant behaviour with reliable physical models and numerical methods adopted for the description of key in-vessel phenomena.

  13. Multi-dimensional boron transport modeling in subchannel approach: Part I. Model selection, implementation and verification of COBRA-TF boron tracking model

    Energy Technology Data Exchange (ETDEWEB)

    Ozdemir, Ozkan Emre, E-mail: ozdemir@psu.edu [Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Avramova, Maria N., E-mail: mna109@psu.edu [Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Sato, Kenya, E-mail: kenya_sato@mhi.co.jp [Mitsubishi Heavy Industries (MHI), Kobe (Japan)

    2014-10-15

    Highlights: ► Implementation of multidimensional boron transport model in a subchannel approach. ► Studies on cross flow mechanism, heat transfer and lateral pressure drop effects. ► Verification of the implemented model via code-to-code comparison with CFD code. - Abstract: The risk of reflux condensation especially during a Small Break Loss Of Coolant Accident (SB-LOCA) and the complications of tracking the boron concentration experimentally inside the primary coolant system have stimulated and subsequently have been a focus of many computational studies on boron tracking simulations in nuclear reactors. This paper presents the development and implementation of a multidimensional boron transport model with Modified Godunov Scheme within a thermal-hydraulic code based on a subchannel approach. The cross flow mechanism in multiple-subchannel rod bundle geometry as well as the heat transfer and lateral pressure drop effects are considered in the performed studies on simulations of deboration and boration cases. The Pennsylvania State University (PSU) version of the COBRA-TF (CTF) code was chosen for the implementation of three different boron tracking models: First Order Accurate Upwind Difference Scheme, Second Order Accurate Godunov Scheme, and Modified Godunov Scheme. Based on the performed nodalization sensitivity studies, the Modified Godunov Scheme approach with a physical diffusion term was determined to provide the best solution in terms of precision and accuracy. As a part of the verification and validation activities, a code-to-code comparison was carried out with the STAR-CD computational fluid dynamics (CFD) code and presented here. The objective of this study was two-fold: (1) to verify the accuracy of the newly developed CTF boron tracking model against CFD calculations; and (2) to investigate its numerical advantages as compared to other thermal-hydraulics codes.

  14. Towards a consistent cosmology with supersymmetry and leptogenesis

    Energy Technology Data Exchange (ETDEWEB)

    Hasenkamp, Jasper

    2012-03-15

    We study the cosmological interplay between supersymmetry, thermal leptogenesis as the origin of matter and the Peccei-Quinn mechanism as solution to the strong CP problem. We investigate to what extent the decay of the lightest ordinary supersymmetric particle, which usually spoils primordial nucleosynthesis in scenarios with gravitino dark matter, can become harmless due to entropy production. We study this possibility for a general neutralino. We find that strong constraints on the entropy-producing particle exclude generic thermal relics as source of sufficient entropy. However, the Peccei-Quinn supermultiplet (axion, saxion, axino) may not only be part of the particle spectrum, but the saxion can also produce a suitable amount of entropy. Exploiting cosmological perturbation theory we show that the corresponding expansion history can be falsified by future observations of the gravitational wave background from inflation, if polarisation measurements of the cosmic microwave background are combined with very sensitive gravitational wave probes. Since the same problem can also be solved by a small breaking of R-parity, we investigate the impact of broken R-parity on the Peccei-Quinn supermultiplet. We find that naturally expected spectra become allowed. Bounds from late particle decays become weaker than those from non-thermal axion production. Thus the strong CP problem serves as an additional motivation for broken R-parity. We show that, if the gravitino problem is solved by a light axino, ''dark radiation'' emerges naturally after primordial nucleosynthesis but before photon decoupling. Current observations of the cosmic microwave background could confirm an increase in the radiation energy density. The other way around, this solution to the gravitino problem implies that thermal leptogenesis might predict such an increase. The Large Hadron Collider could endorse this opportunity. In the same parameter range, axion and axino can naturally

  15. Light-Weight Free-Standing Carbon Nanotube-Silicon Films for Anodes of Lithium Ion Batteries

    KAUST Repository

    Cui, Li-Feng

    2010-07-27

    Silicon is an attractive alloy-type anode material because of its highest known capacity (4200 mAh/g). However, lithium insertion into and extraction from silicon are accompanied by a huge volume change, up to 300%, which induces a strong strain on silicon and causes pulverization and rapid capacity fading due to the loss of the electrical contact between part of silicon and current collector. Si nanostructures such as nanowires, which are chemically and electrically bonded to the current collector, can overcome the pulverization problem, however, the heavy metal current collectors in these systems are larger in weight than Si active material. Herein we report a novel anode structure free of heavy metal current collectors by integrating a flexible, conductive carbon nanotube (CNT) network into a Si anode. The composite film is free-standing and has a structure similar to the steel bar reinforced concrete, where the infiltrated CNT network functions as both mechanical support and electrical conductor and Si as a high capacity anode material for Li-ion battery. Such free-standing film has a low sheet resistance of ∼30 Ohm/sq. It shows a high specific charge storage capacity (∼2000 mAh/g) and a good cycling life, superior to pure sputtered-on silicon films with similar thicknesses. Scanning electron micrographs show that Si is still connected by the CNT network even when small breaking or cracks appear in the film after cycling. The film can also "ripple up" to release the strain of a large volume change during lithium intercalation. The conductive composite film can function as both anode active material and current collector. It offers ∼10 times improvement in specific capacity compared with widely used graphite/copper anode sheets. © 2010 American Chemical Society.

  16. Calculation of Noninformative Prior of Reliability Parameter and Initiating Event Frequency With Jeffreys Method%应用Jeffreys方法计算可靠性参数和始发事件频率的无信息先验

    Institute of Scientific and Technical Information of China (English)

    何劼; 张彬彬

    2013-01-01

    在核电厂概率安全评价(PSA )分析中,有些始发事件频率或设备失效记录在工业界几乎无历史数据。为了计算这些无信息先验的可靠性参数和始发事件频率,可采用Bayesian统计学中的Jeffreys方法。本文阐述了Jeffreys先验和简化的受限无信息先验分布(SCNID )的数学原理,分别导出了Gamma-Poisson模型和Beta-Binomial模型的Jeffreys无信息先验公式和不确定性区间。结合反应堆冷却剂小破口失水事故(SLOCA)实例介绍了如何应用Jeffreys先验计算始发事件频率。结果表明,Jeffreys方法是一种计算无信息先验的有效方法。%In the probabilistic safety assessment (PSA ) of nuclear power plants ,there are few historical records on some initiating event frequencies or component failures in industry .In order to determine the noninformative priors of such reliability parameters and initiating event frequencies , the Jeffreys method in Bayesian statistics was employed . The mathematical mechanism of the Jeffreys prior and the simplified constrained noninformative distribution (SCNID) were elaborated in this paper .The Jeffreys noninformative formulas and the credible intervals of the Gamma-Poisson and Beta-Binomial models were introduced .As an example ,the small break loss-of-coolant accident (SLOCA) was employed to show the application of the Jeffreys prior in deter-mining an initiating event frequency .The result shows that the Jeffreys method is an effective method for noninformative prior calculation .

  17. Gene Flow of a Forest-Dependent Bird across a Fragmented Landscape.

    Directory of Open Access Journals (Sweden)

    Rachael V Adams

    Full Text Available Habitat loss and fragmentation can affect the persistence of populations by reducing connectivity and restricting the ability of individuals to disperse across landscapes. Dispersal corridors promote population connectivity and therefore play important roles in maintaining gene flow in natural populations inhabiting fragmented landscapes. In the prairies, forests are restricted to riparian areas along river systems which act as important dispersal corridors for forest dependent species across large expanses of unsuitable grassland habitat. However, natural and anthropogenic barriers within riparian systems have fragmented these forested habitats. In this study, we used microsatellite markers to assess the fine-scale genetic structure of a forest-dependent species, the black-capped chickadee (Poecile atricapillus, along 10 different river systems in Southern Alberta. Using a landscape genetic approach, landscape features (e.g., land cover were found to have a significant effect on patterns of genetic differentiation. Populations are genetically structured as a result of natural breaks in continuous habitat at small spatial scales, but the artificial barriers we tested do not appear to restrict gene flow. Dispersal between rivers is impeded by grasslands, evident from isolation of nearby populations (~ 50 km apart, but also within river systems by large treeless canyons (>100 km. Significant population genetic differentiation within some rivers corresponded with zones of different cottonwood (riparian poplar tree species and their hybrids. This study illustrates the importance of considering the impacts of habitat fragmentation at small spatial scales as well as other ecological processes to gain a better understanding of how organisms respond to their environmental connectivity. Here, even in a common and widespread songbird with high dispersal potential, small breaks in continuous habitats strongly influenced the spatial patterns of genetic

  18. Gene Flow of a Forest-Dependent Bird across a Fragmented Landscape

    Science.gov (United States)

    2015-01-01

    Habitat loss and fragmentation can affect the persistence of populations by reducing connectivity and restricting the ability of individuals to disperse across landscapes. Dispersal corridors promote population connectivity and therefore play important roles in maintaining gene flow in natural populations inhabiting fragmented landscapes. In the prairies, forests are restricted to riparian areas along river systems which act as important dispersal corridors for forest dependent species across large expanses of unsuitable grassland habitat. However, natural and anthropogenic barriers within riparian systems have fragmented these forested habitats. In this study, we used microsatellite markers to assess the fine-scale genetic structure of a forest-dependent species, the black-capped chickadee (Poecile atricapillus), along 10 different river systems in Southern Alberta. Using a landscape genetic approach, landscape features (e.g., land cover) were found to have a significant effect on patterns of genetic differentiation. Populations are genetically structured as a result of natural breaks in continuous habitat at small spatial scales, but the artificial barriers we tested do not appear to restrict gene flow. Dispersal between rivers is impeded by grasslands, evident from isolation of nearby populations (~ 50 km apart), but also within river systems by large treeless canyons (>100 km). Significant population genetic differentiation within some rivers corresponded with zones of different cottonwood (riparian poplar) tree species and their hybrids. This study illustrates the importance of considering the impacts of habitat fragmentation at small spatial scales as well as other ecological processes to gain a better understanding of how organisms respond to their environmental connectivity. Here, even in a common and widespread songbird with high dispersal potential, small breaks in continuous habitats strongly influenced the spatial patterns of genetic variation. PMID

  19. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    Science.gov (United States)

    Scarlat, Raluca Olga

    local buoyancy effects Experiments indicate that slow exchange of stagnant fluid in static legs can play a significant role in the transient response of natural circulation loops. The effect of non-linear temperature profiles on the hot or cold legs or other segments of the flow loop, which may develop during transient scenarios, should be considered when modeling the performance of natural circulation loops. The data provided here can be used for validation of the application of thermal-hydraulic systems codes to the modeling of heat removal by natural circulation with liquid fluoride salts and its simulant fluids.

  20. 反应堆一回路系统优化设计方案的可行性验证%Feasibility Test for Reactor Coolant System Optimized Design Scheme

    Institute of Scientific and Technical Information of China (English)

    陈磊; 阎昌琪; 王建军

    2014-01-01

    采用优选运行参数和结构参数的方法,可达到降低核动力装置尺寸的目的。在优化设计方案投入制造前,有必要研究其在设计基准事故下的响应特性,以检验优化方案的可行性。采用 REL A P5/M OD3.2程序研究现有一回路系统优化方案在完全失去厂外电、主给水丧失和小破口失水事故下的响应特性,并将安全设计准则参数与母型对比。结果表明:针对所研究的3种设计基准事故,优化方案各主要安全准则参数满足设计要求;优化方案可成功抵御这3类设计基准事故。%The size of a nuclear component could be reduced by optimum selections of the operational and structural parameters .Before an optimized design scheme is manu‐factured ,it is necessary to obtain its transient behaviors and verify its feasibility under design basis accidents .In this work ,the RELAP5/MOD3.2 code was employed to simulate the transient characteristics of a proposed optimized scheme under the complete loss of off‐site power ,loss of feedwater and small break loss of coolant accidents ,and the safety criteria were compared with the prototype reactor design . The simulation results indicate that the safety criteria of the optimized scheme satisfy the design requirements ,and the safety of the optimized scheme can be guaranteed in those three accidents .

  1. Effect of In-Vessel Retention Strategies under Postulated SGTR Accidents of OPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Wonjun; Lee, Yongjae; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of); Kim, Hwan-Yeol; Park, Rae-Joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, MELCOR code was used to simulate the severe accident of the OPR1000. MELCOR code is computer code which enables to simulate the progression of the severe accident for light water reactors. It has been developed by Sandia National Laboratories for plant risk assessment and source term analysis since 1982. According to the probabilistic safety analysis (PSA) Level 1 of OPR1000, typical severe accident scenarios of high probability of a transition to severe accident for OPR1000 were identified as Small Break Loss of Coolant Accident (SBLOCA), Station Black out (SBO), Total Loss of Feed Water (TLOFW), and Steam Generator Tube Rupture. While the first three accidents are expected to result in the generation and transportation of the radioactive nuclides within the containment building as consequence of the core damage and subsequent reactor pressure vessel (RPV) failure, the latter accident scenario may be progressed with possible direct release of the radioactive nuclides to the environment by bypassing the containment building. Thus it is of significance to investigate the SGTR accident with a sophisticated severe accident code. This code can simulate the whole phenomena of a severe accident such as thermal-hydraulic response, core heat-up, oxidation and relocation, and fission product release and transport. Thus many researchers have used MELCOR in severe accident studies. In this study, in-vessel retention strategies were applied for postulated SGTR accidents. Mitigation effect and adverse effect of in-vessel strategies was studied in aspect of RPV failure, fission product release and containment thermal-hydraulic and hydrogen behavior. Base case of SGTR accident and three mitigation cases were simulated using MELCOR code 1.8.6. For each mitigation cases, mitigation effect and adverse effect were investigated. Conclusions can be summarized as follows: (1) RPV failure of SGTR base case occurred at 5.62 hours and fission product of RCS released to

  2. ThermalGhydraulic Simulation of DEDVI Accident for Advanced Passive PWR%先进非能动核电厂DEDVI事故热工水力模拟分析

    Institute of Scientific and Technical Information of China (English)

    余健明; 曹学武

    2016-01-01

    The accident analysis model is established by the code of Relap5/Mod 3.4, which includes the Reactor Coolant System (RCS),simplified secondary system and Engineering Safety Features (ESF). A typical Small-Break LOCA(SBLOCA)accident, Double-Ended Direct Vessel Inj ection (DEDVI ), is selected to analyze the accident scenario and sensitivity analyses of entrainment models have been taken with respect to pressure,mass flow rate,liquid levels and peak cladding temperature. The results show that the break and ADS system can depressurize the RCS quickly and the coolant from CMT,ACC and IRWST can mitigate the accidental consequence of DEDVI effectively. Sensitivity analysis of entrainment models shows that homogenous flow model creates higher liquid discharge flow rate comparing to nonhomogenous flow model.%采用 Relap5/Mod3.4程序建立了先进非能动核电厂的事故分析模型,包括反应堆冷却剂系统(RCS)、简化的二回路系统和专设安全设施.针对小破口失水事故(SBLOCA)中的直接安注管双端断裂事故(DEDVI)进行分析,并着重对 SBLOCA 现象识别和排序表(PIRT)中对其影响较大的液滴夹带进行敏感性分析.分析结果表明,对直接安注管双端断裂事故,破口和自动卸压系统(ADS)能够有效地使反应堆冷却剂系统降压,堆芯补水箱(CMT)、安注箱(ACC)和安全壳内置换料水箱(IRWST)能够迅速实现堆芯补水,确保堆芯冷却.对液滴夹带的敏感性分析表明,对于位置较高的第4级 ADS,喷放流量对液滴夹带模型比较敏感,使用均相流模型计算时,其液相流量显著高于非均相流模型.

  3. Development of a feed-and-bleed operation strategy with hybrid-SIT under low pressure condition of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, In Seop, E-mail: jeoni@rpi.edu [Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, 110 8th Street, Troy, NY (United States); Han, Sang Hoon, E-mail: shhan2@kaeri.re.kr [Advanced Research Group, Korea Atomic Energy Research Institute, 70 Daedeok-daero 989 Beon-gil, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Kang, Sang Hee, E-mail: sanghee.kang@khnp.co.kr [NSSS Design Group, Korea Hydro & Nuclear Power Co., Ltd., Central Research Institute, 70, 1312-beongil, Yuseongdaero, Yuseong-gu, Daejeon (Korea, Republic of); Kang, Hyun Gook, E-mail: hyungook@kaist.ac.kr [Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, 110 8th Street, Troy, NY (United States)

    2017-04-01

    plant by 5.0 percent in the case of a small break loss of coolant accident. Total core damage frequency of the plant decreases by 4.8 percent compared to the reference model, and also lowers the number of cutsets by around 13 percent from the original value.

  4. Establishment and assessment of code scaling capability

    Science.gov (United States)

    Lim, Jaehyok

    In this thesis, a method for using RELAP5/MOD3.3 (Patch03) code models is described to establish and assess the code scaling capability and to corroborate the scaling methodology that has been used in the design of the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility. It was sponsored by the United States Nuclear Regulatory Commission (USNRC) under the program "PUMA ESBWR Tests". PUMA-E facility was built for the USNRC to obtain data on the performance of the passive safety systems of the General Electric (GE) Nuclear Energy Economic Simplified Boiling Water Reactor (ESBWR). Similarities between the prototype plant and the scaled-down test facility were investigated for a Gravity-Driven Cooling System (GDCS) Drain Line Break (GDLB). This thesis presents the results of the GDLB test, i.e., the GDLB test with one Isolation Condenser System (ICS) unit disabled. The test is a hypothetical multi-failure small break loss of coolant (SB LOCA) accident scenario in the ESBWR. The test results indicated that the blow-down phase, Automatic Depressurization System (ADS) actuation, and GDCS injection processes occurred as expected. The GDCS as an emergency core cooling system provided adequate supply of water to keep the Reactor Pressure Vessel (RPV) coolant level well above the Top of Active Fuel (TAF) during the entire GDLB transient. The long-term cooling phase, which is governed by the Passive Containment Cooling System (PCCS) condensation, kept the reactor containment system that is composed of Drywell (DW) and Wetwell (WW) below the design pressure of 414 kPa (60 psia). In addition, the ICS continued participating in heat removal during the long-term cooling phase. A general Code Scaling, Applicability, and Uncertainty (CSAU) evaluation approach was discussed in detail relative to safety analyses of Light Water Reactor (LWR). The major components of the CSAU methodology that were highlighted particularly focused on the

  5. Preliminary assessment of a combined passive safety system for typical 3-loop PWR CPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Zijiang; Shan, Jianqiang, E-mail: jqshan@mail.xjtu.edu.cn; Gou, Junli

    2017-03-15

    Highlights: • A combined passive safety system was placed on a typical 3-loop PWR CPR1000. • Three accident analyses show the three different accident mitigation methods of the passive safety system. • The three mitigation methods were proved to be useful. - Abstract: As the development of the nuclear industry, passive technology turns out to be a remarkable characteristic of advanced nuclear power plants. Since the 20th century, much effort has been given to the passive technology, and a number of evolutionary passive systems have developed. Thoughts have been given to upgrade the existing reactors with passive systems to meet stricter safety demands. In this paper, the CPR1000 plant, which is one kind of mature pressurized water reactor plants in China, is improved with some passive systems to enhance safety. The passive systems selected are as follows: (1) the reactor makeup tank (RMT); (2) the advanced accumulator (A-ACC); (3) the in-containment refueling water storage tank (IRWST); (4) the passive emergency feed water system (PEFS), which is installed on the secondary side of SGs; (5) the passive depressurization system (PDS). Although these passive components is based on the passive technology of some advanced reactors, their structural and trip designs are adjusted specifically so that it could be able to mitigate accidents of the CPR1000. Utilizing the RELAP5/MOD3.3 code, accident analyses (small break loss of coolant accident, large break loss of coolant accident, main feed water line break accident) of this improved CPR1000 plant were presented to demonstrate three different accident mitigation methods of the safety system and to test whether the passive safety system preformed its function well. In the SBLOCA, all components of the passive safety system were put into work sequentially, which prevented the core uncover. The LBLOCA analysis illustrates the contribution of the A-ACCs whose small-flow-rate injection can control the maximum cladding

  6. Integration of the functional reliability of two passive safety systems to mitigate a SBLOCA+BO in a CAREM-like reactor PSA

    Energy Technology Data Exchange (ETDEWEB)

    Mezio, Federico, E-mail: federico.mezio@cab.cnea.gov.ar [CNEA, Sede Central, Av. Del Libertador 8250, CABA (Argentina); Grinberg, Mariela [CNEA, Centro Atómico Bariloche, S.C. de Bariloche, Río Negro (Argentina); Lorenzo, Gabriel [CNEA, Sede Central, Av. Del Libertador 8250, CABA (Argentina); Giménez, Marcelo [CNEA, Centro Atómico Bariloche, S.C. de Bariloche, Río Negro (Argentina)

    2014-04-01

    Highlights: • An estimation of the Functional Unreliability was performed using RMPS methodology. • The methodology uses an improved response surface in order to estimate the FU. • The FU may become relevant to be analyzed in the Passive Safety Systems. • There were proposed two ways to incorporate the FU into an APS. - Abstract: This paper describes a case study of a methodological approach for assessing the functional reliability of passive safety systems (PSS) and its treatment within a probabilistic safety assessment (PSA). The functional unreliability (FU) can be understood as the failure probability of PSS to fulfill its mission due to the impairment of the related passive safety function. The safety function accomplishment is characterized and quantified by a performance indicator (PI), which is a measure of how far the system is from verifying its mission. PI uncertainties are estimated from uncertainty propagation of selected parameters. A methodology based on the reliability methodology for passive system (RMPS) one is used to estimate the FU associated to the isolation condensers (ICs) in combination with the accumulators (medium pressure injection system) of a CAREM-like integral advanced reactor. A small break loss of coolant accident with black-out is selected as an evaluation case. This implies success of reactor shut-down (inherent) and failure of residual heat removal by active systems. The safety function to accomplish is to refill the reactor pressure vessel (RPV) in order to avoid core damage. For this case, to allow the discharge of accumulators into RPV, the pressure must be reduced by the IC. The methodology for passive safety function assessment considers uncertainties in code parameters, besides uncertainties in engineering parameters (design, construction, operation and maintenance), in order to perform Monte Carlo simulations based on best estimate (B-E) plant model. Then, response surfaces based on PI are used for improving the

  7. Reactor Core Coolability Analysis during Hypothesized Severe Accidents of OPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yongjae; Seo, Seungwon; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of); Ha, Kwang Soon; Kim, Hwan-Yeol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Assessment of the safety features over the hypothesized severe accidents may be performed experimentally or numerically. Due to the considerable time and expenditures, experimental assessment is implemented only to the limited cases. Therefore numerical assessment has played a major role in revisiting severe accident analysis of the existing or newly designed power plants. Computer codes for the numerical analysis of severe accidents are categorized as the fast running integral code and detailed code. Fast running integral codes are characterized by a well-balanced combination of detailed and simplified models for the simulation of the relevant phenomena within an NPP in the case of a severe accident. MAAP, MELCOR and ASTEC belong to the examples of fast running integral codes. Detailed code is to model as far as possible all relevant phenomena in detail by mechanistic models. The examples of detailed code is SCDAP/RELAP5. Using the MELCOR, Carbajo. investigated sensitivity studies of Station Black Out (SBO) using the MELCOR for Peach Bottom BWR. Park et al. conduct regulatory research of the PWR severe accident. Ahn et al. research sensitivity analysis of the severe accident for APR1400 with MELCOR 1.8.4. Lee et al. investigated RCS depressurization strategy and developed a core coolability map for independent scenarios of Small Break Loss-of-Coolant Accident (SBLOCA), SBO, and Total Loss of Feed Water (TLOFW). In this study, three initiating cases were selected, which are SBLOCA without SI, SBO, and TLOFW. The initiating cases exhibit the highest probability of transitioning into core damage according to PSA 1 of OPR 1000. The objective of this study is to investigate the reactor core coolability during hypothesized severe accidents of OPR1000. As a representative indicator, we have employed Jakob number and developed JaCET and JaMCT using the MELCOR simulation. Although the RCS pressures for the respective accident scenarios were different, the JaMCT and Ja

  8. Use of Structure-from-Motion Photogrammetry Technique to model Danxia red bed landform slope stability by discrete element modeling - case study at Mt. Langshan, Hunan Province, China

    Science.gov (United States)

    Simonson, Scott; Hua, Peng; Luobin, Yan; Zhi, Chen

    2016-04-01

    Important to the evolution of Danxia landforms is how the rock cliffs are in large part shaped by rock collapse events, ranging from small break offs to large collapses. Quantitative research of Danxia landform evolution is still relatively young. In 2013-2014, Chinese and Slovak researchers conducted joint research to measure deformation of two large rock walls. In situ measurements of one rock wall found it to be stable, and Ps-InSAR measurements of the other were too few to be validated. Research conducted this year by Chinese researchers modeled the stress states of a stone pillar at Mt. Langshan, in Hunan Province, that toppled over in 2009. The model was able to demonstrate how stress states within the pillar changed as the soft basal layer retreated, but was not able to show the stress states at the point of complete collapse. According to field observations, the back side of the pillar fell away from the entire cliff mass before the complete collapse, and no models have been able to demonstrate the mechanisms behind this behavior. A further understanding of the mechanisms controlling rockfall events in Danxia landforms is extremely important because these stunning sceneries draw millions of tourists each year. Protecting the tourists and the infrastructure constructed to accommodate tourism is of utmost concern. This research will employ a UAV to as universally as possible photograph a stone pillar at Mt. Langshan that stands next to where the stone pillar collapsed in 2009. Using the recently developed structure-from-motion technique, a 3D model of the pillar will be constructed in order to extract geometrical data of the entire slope and its structural fabric. Also in situ measurements will be taken of the slope's toe during the field work exercises. These data are essential to constructing a realistic discrete element model using the 3DEC code and perform a kinematic analysis of the rock mass. Intact rock behavior will be based on the Mohr Coulomb

  9. Investigation of the melt-down behaviour of massive radial core enclosures during LWR accidents

    Energy Technology Data Exchange (ETDEWEB)

    Hering, W.; Sengpiel, W.; Messainguiral, C. [CEA Centre d' Etudes Nucleaires de Cadarache, 13 - Saint-Paul-les-Durance (France). DRN

    2000-11-01

    At the Institute for Reactor Safety (IRS) of the Forschungszentrum Karlsruhe (FZK) accident analyses were performed for the projected European pressurised water reactor (EPR) up to 1999 using the best estimate severe core damage code SCDAP/RELAP5 (S/R5). From various scenarios investigated with S/R5 the loss-of-offsite power (LOOP) and the 46 cm{sup 2} small break loss of coolant accident (SBLOCA) were selected to be discussed here in some detail. To simulate the heavy reflector (HR) and core barrel (CB) behaviour beyond the capabilities of S/R5 mod 3.2 a detailed stand alone analytical tool (LOWCOR2) was developed and used to determine the time of HR melting, its axial position, the melting velocity and the melt mass. Furthermore, results of MELCOR calculations performed at Siemens/KWU were used for the SBLOCA scenario. The analyses were extended by a feasibility study to find out whether ICARE2 and the commercial FEM code FIDAP are applicable. The axial position of HR and CB melt through strongly depends on the scenario an ranges between 1.0 m and 2.5 m core elevation. The time period to melt down the HR inner edges lasts up to 17 min and a complete melt through of HR and CB is in the order of magnitude of one hour. At melt through time LOWCOR2 calculated a molten steel mass between 10 Mg and 32 Mg and a melt relocation rate of 35 kg/s along the HR inner surface into the core cavity. (orig.) [German] Am Institut fuer Reaktorsicherheit (IRS) des Forschungszentrums Karlsruhe (FZK) wurden bis 1999 Unfallanalysen fuer den projektierten Europaeischen Druckwasser Reaktor (EPR) mit dem ''best estimate'' Kernschmelzcode SCDAP/RELAP5 (S/R5) durchgefuehrt. Von den verschiedenen mit S/R5 untersuchten Szenarien wurden der ''Ausfall der Wechselspannungsnetze'' (LOOP) und das kleine Leck (46 cm{sup 2}) im kalten Strang der Hauptkuehlmittelleitung (SBLOCA) fuer eine ausfuehrlichere Diskussion ausgewaehlt. Um das Verhalten des &apos

  10. The ALOMAR Rayleigh/Mie/Raman lidar: objectives, configuration, and performance

    Science.gov (United States)

    von Zahn, U.; von Cossart, G.; Fiedler, J.; Fricke, K. H.; Nelke, G.; Baumgarten, G.; Rees, D.; Hauchecorne, A.; Adolfsen, K.

    2000-07-01

    to 36 detector channels simultaneously record the photons collected by the telescopes. The internal and external instrument operations are automated so that this very complex instrument can be operated by a single engineer. Currently the lidar is heavily used for measurements of temperature profiles, of cloud particle properties such as their altitude, particle densities and size distributions, and of stratospheric winds. Due to its very effective spectral and spatial filtering, the lidar has unique capabilities to work in full sunlight. Under these conditions it can measure temperatures up to 65 km altitude and determine particle size distributions of overhead noctilucent clouds. Due to its very high mechanical and optical stability, it can also employed efficiently under marginal weather conditions when data on the middle atmosphere can be collected only through small breaks in the tropospheric cloud layers.

  11. The ALOMAR Rayleigh/Mie/Raman lidar: objectives, configuration, and performance

    Directory of Open Access Journals (Sweden)

    U. von Zahn

    >+O2 molecules, respectively. Currently, up to 36 detector channels simultaneously record the photons collected by the telescopes. The internal and external instrument operations are automated so that this very complex instrument can be operated by a single engineer. Currently the lidar is heavily used for measurements of temperature profiles, of cloud particle properties such as their altitude, particle densities and size distributions, and of stratospheric winds. Due to its very effective spectral and spatial filtering, the lidar has unique capabilities to work in full sunlight. Under these conditions it can measure temperatures up to 65 km altitude and determine particle size distributions of overhead noctilucent clouds. Due to its very high mechanical and optical stability, it can also employed efficiently under marginal weather conditions when data on the middle atmosphere can be collected only through small breaks in the tropospheric cloud layers.

    Key words: Atmospheric composition and structure (aerosols and particles; pressure · density · and temperature; instruments and techniques