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Journal Articles

RELAP5 analysis of OECD/NEA ROSA project experiment simulating a PWR loss-of-feedwater transient with high-power natural circulation

Takeda, Takeshi; Asaka, Hideaki*; Nakamura, Hideo

Science and Technology of Nuclear Installations, 2012, p.957285_1 - 957285_15, 2012/00

 Times Cited Count:10 Percentile:31.32(Nuclear Science & Technology)

A ROSA/LSTF experiment was conducted for OECD/NEA ROSA Project simulating a PWR loss-of-feedwater transient with assumptions of failure of scram and total failure of HPI system. AFW was provided to well observe the long-term high-power natural circulation. The core power curve was obtained from a RELAP5 code analysis of PWR LOFW transient without scram. The primary and SG secondary-side pressures were maintained respectively at around 16 and 8 MPa by cycle opening of PZR PORV and SG relief valves. Large-amplitude level oscillation occurred in SG U-tubes for a long time in a form of slow fill and dump while the two-phase natural circulation flow rate gradually decreased with some oscillation. RELAP5 post-test analyses were performed by employing a fine-mesh multiple parallel flow channel representation of SG U-tubes with a Wallis CCFL correlation. Problems remain in the predictions of the oscillative primary loop flow rate and SG U-tube liquid level as well as PZR liquid level.

Journal Articles

RELAP5 post-test analyses of OECD/NEA ROSA project experiments on steam generator depressurization with or without non-condensable gas inflow

Takeda, Takeshi; Asaka, Hideaki*; Watanabe, Tadashi; Nakamura, Hideo

Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 8 Pages, 2008/11

Two LSTF experiments were conducted for OECD/NEA ROSA Project simulating PWR 0.5% cold leg small break LOCA. Steam generator (SG) secondary-side depressurization was performed by fully opening the relief valves at 10 minutes after safety injection signal with or without non-condensable gas (air) inflow from accumulator tanks with total failure of high pressure injection system. Further assumptions were made to conduct enhanced SG depressurization by fully opening the safety valves when the primary pressure decreased to 2 MPa and no actuation of low pressure injection system, both to well observe natural circulation (NC) phenomena at low pressures. The primary depressurization rate decreased when non-condensable gas started to enter primary loops because of degradation in the condensation heat transfer in SG U-tubes, while two-phase flow NC has continued even after non-condensable gas inflow. Asymmetric NC behaviors appeared between two loops due probably to different number of forward flow SG U-tubes which would have been under influences of non-condensable gas. Post-test analyses by using JAEA-modified RELAP5/MOD3.2.1.2 code indicated that the code has remaining problems in proper prediction of primary loop flow rate and SG U-tube liquid level behaviors especially after non-condensable gas inflow. The improvement of the condensation heat transfer model under non-condensable gas mixture condition and the SG U-tube model may be necessary for correct analysis of the LSTF SG depressurization transients.

Journal Articles

RELAP5 analysis of OECD/NEA ROSA Project experiment simulating a PWR small break LOCA with high-power natural circulation

Takeda, Takeshi; Asaka, Hideaki*; Nakamura, Hideo

Proceedings of International Conference on the Physics of Reactors, Nuclear Power; A Sustainable Resource (PHYSOR 2008) (CD-ROM), 8 Pages, 2008/09

OECD/NEA ROSA Project experiment with the Large Scale Test Facility (LSTF) in JAEA was conducted simulating a PWR 1% cold leg small break LOCA with an assumption of high-power natural circulation due to failure of scram and total failure of high pressure injection system. The core power curve for the LSTF experiment was obtained by PWR analysis with coupled three-dimensional kinetics and thermal-hydraulics code SKETCH-INS/TRAC-PF1. A post-test analysis was performed against the obtained data by using JAEA-modified RELAP5/MOD3.2.1.2 code to validate the code predictability. The JAEA-modified RELAP5/MOD3.2.1.2 code was used by incorporating a break model that employs maximum bounding flow theory with a discharge coefficient of 0.61 for two-phase break flow. In the experiment, flow in hot legs became supercritical during two-phase natural circulation, causing the hot leg liquid level to be quite low. Liquid accumulation in steam generator U-tube upflow-side took place during reflux condensation mode. The RELAP5 code predicted reasonably well the overall thermal-hydraulic phenomena observed in the experiment, while it overpredicted the break flow rate.

Journal Articles

RELAP5 analysis of ROSA/LSTF vessel upper head break LOCA experiment

Takeda, Takeshi; Asaka, Hideaki; Suzuki, Mitsuhiro; Nakamura, Hideo

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12) (CD-ROM), 3 Pages, 2007/09

RELAP5 code analysis was performed to validate the code predictability by using ROSA/LSTF experiment data that simulated a PWR vessel upper head small break loss-of-coolant accident (SBLOCA) with a break equivalent to 1% cold leg break. The JAEA-modified RELAP5/MOD3.2.1.2 code was used by incorporating a break model that employs maximum bounding flow theory with a discharge coefficient (Cd) of 0.61 for two-phase break flow. In the experiment, liquid level in the upper head was found to control break flow rate as coolant in the upper plenum entered the upper head through control rod guide tubes (CRGTs) until the penetration holes at the CRGT bottom were exposed to steam in the upper plenum. The upper head noding and flow paths between the upper plenum and the CRGT were thus modeled to simulate well the liquid level and coolant flow around the upper portion of pressure vessel. The code, however, overpredicted the break flow rate due to the underprediction of break-upstream void fraction especially during two-phase flow discharge period. Cd for two-phase break flow was thus adjusted to be 0.58. Effects of break area on the core cooling were investigated further. The parameter analyses showed that peak cladding temperature (PCT) is the maximum at 1% break case, while the PCT would be lower than 1200 K in the larger break size cases because vapor condensation on injected accumulator coolant induces loop seal clearing and effectively enhances core cooling thereafter.

JAEA Reports

A Study on timing of rapid depressurization action during PWR vessel bottom break LOCA with HPI failure and AIS-gas inflow, ROSA-V/LSTF test SB-PV-06

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

JAEA-Research 2007-037, 150 Pages, 2007/03

JAEA-Research-2007-037.pdf:7.55MB

A small break LOCA experiment (SB-PV-06) was conducted at the LSTF of ROSA-V program to study effects of rapid secondary depressuriza-tion action on core cooling as one of accident management (AM) measures for a PWR in case of high pressure injection system failure and non-condensable gas inflow from the accumulator injection system. The break simulated 10 instrument tubes rupture equivalent to 0.2% cold leg break. It was clarified through comparison with former experiments that (1) the depressurization initiated by detecting the vessel level below the primary loop (4545s) was degraded by the gas inflow resulting in whole core uncovery prior to the start of low pressure injection and (2) an alternative start of the depressurization by detecting level decrease at the SG outlet plenum (2330s), would limit the core uncovery suggesting more effective parameter for the AM measures. The report presents the experiment results with the effects of rapid depressurization initiation timing.

Journal Articles

Study on transient void behavior during reactivity initiated accidents under low pressure condition; Development and application of measurement technique for void fraction in bundle geometry

Satou, Akira; Maruyama, Yu; Asaka, Hideaki; Nakamura, Hideo

Journal of Power and Energy Systems (Internet), 1(2), p.154 - 165, 2007/00

Series of out-of-pile experiments to obtain the knowledge on the transient void behavior during reactivity initiated accidents are in progress at JAEA. In the present series of experiments, the transient void behavior in a test section of 2$$times$$2 bundle geometry under atmospheric pressure condition was measured using an impedance technique. The measuring areas and the arrangement of electrodes for the impedance technique were defined on the basis of numerical analyses and scaled model experiments. The comparison was made between the impedance and differential pressure techniques for steady boiling experiments to estimate the accuracy of the impedance technique. The impedance technique showed a good agreement with the void fraction estimated from the differential pressure. The transient void behavior in the bundle geometry was measured using the impedance technique. It was clarified that the transient void behavior depends on both the subcooling of inlet water and the heat generation rate of simulated fuel rod. Local void fraction was influenced by the ratio of flow area to heat transfer area of the simulated fuel rod.

JAEA Reports

A Study on effective system depressurization during a PWR vessel bottom break LOCA with HPI failure and gas inflow prevention (ROSA-V/LSTF test SB-PV-05)

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

JAEA-Research 2006-072, 144 Pages, 2006/11

JAEA-Research-2006-072.pdf:17.16MB

A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection (HPI) system during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating a 4-loop Westing-house-type PWR (3423 MWt). The experiment, SB-PV-05, simulated a PWR vessel bottom SBLOCA with a rupture of nine instrument tubes, which is equivalent to 0.18% cold leg break. It is clarified that AM actions with steam generator (SG) depressurization to achieve a primary loop cooling rate at -55 K/h and auxiliary feedwater supply for 30 minutes are effective to avoid core uncovery by actuating the low pressure injection (LPI) system. It is also shown through the comparison with the previous experiment of SB-PV-03 that prevention of non-condensable gas inflow from the accumulator injection system (AIS) is very important to actuate the LPI to achieve adequate core cooling. This report presents experiment results of SB-PV-05 in detail and shows the effects of gas inflow prevention on core cooling through the estimation of primary coolant mass and energy balance in the primary system.

Journal Articles

Study on transient void behavior during reactivity initiated accidents under low pressure condition; Development and application of measurement technique for void fraction in bundle geometry

Satou, Akira; Maruyama, Yu; Asaka, Hideaki; Nakamura, Hideo

Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 8 Pages, 2006/07

Series of out-of-pile experiments to obtain the knowledge on the transient void behavior during reactivity initiated accidents are in progress at JAEA. In the present series of experiments, the transient void behavior in a test section of 2$$times$$2 bundle geometry under atmospheric pressure condition was measured using an impedance technique. The measuring areas and the arrangement of electrodes for the impedance technique were defined on the basis of numerical analyses and scaled model experiments. The comparison was made between the impedance and differential pressure techniques for steady boiling experiments to estimate the accuracy of the impedance technique. The impedance technique showed a good agreement with the void fraction estimated from the differential pressure. The transient void behavior in the bundle geometry was measured using the impedance technique. It was clarified that the transient void behavior depends on both the subcooling of inlet water and the heat generation rate of simulated fuel rod. Local void fraction was influenced by the ratio of flow area to heat transfer area of the simulated fuel rod.

JAEA Reports

An Experimental study on effective depressurization actions for PWR vessel bottom small break LOCA with HPI failure and gas inflow, ROSA-V test SB-PV-04

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

JAEA-Research 2006-018, 140 Pages, 2006/03

JAEA-Research-2006-018.pdf:7.14MB

A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection system during an SBLOCA at a pressurized water reactor (PWR). The experiment (SB-PV-04) simulating a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes clarified that rapid depressurization action by full-opening of relief valves and supplying auxiliary feedwater were effective to avoid core uncovery through actuation of low pressure injection system irrespective of significantly degraded depressurization by non-condensable gas inflow from the accumulator tanks. It is clarified that the effective core cooling was established by the rapid primary cooling which contributed to preserve larger primary coolant mass than in the previous experiment (SB-PV-03) which was conducted with smaller primary cooling rate of -55 K/h as another AM action and resulted in core heatup.

Journal Articles

Effects of secondary depressurization on core cooling in PWR vessel bottom small break LOCA experiments with HPI failure and gas inflow

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

Journal of Nuclear Science and Technology, 43(1), p.55 - 64, 2006/01

 Times Cited Count:10 Percentile:38.8(Nuclear Science & Technology)

Effects of non-condensable gas from the accumulator tanks on secondary depressurization, as one of accident management (AM) measures in case of high pressure injection system failure, are studied at the ROSA-V/LSTF experiments simulating a ten instrument-tube break LOCA at the PWR vessel bottom. In an experiment with no gas inflow, the secondary depressurization at -55 K/h initiated by SI signal with 10 minutes delay succeeded in the LPI actuation. On the other hand, the gas inflow in another experiment degraded the primary depressurization and resulted in core uncovery before the LPI start. The third experiment with rapid secondary depressurization and continuous auxiliary feedwater supply, however, showed a possibility of long-term core cooling by the LPI actuation. RELAP5/MOD3 code analyses well predicted these experiment results and clarified that condensation heat transfer was largely degraded by the gas in the U-tubes. In addition, a primary pressure - coolant mass map was found to be useful for indication of key plant parameters of AM measures.

Journal Articles

A ROSA-V experimental study on PWR accident management actions and role of reactor instrumentation

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

Nippon Kikai Gakkai 2005-Nendo Nenji Taikai Koen Rombunshu, Vol.3, p.223 - 224, 2005/09

Shown below are experimental results on characteristics of reactor instrumentations including a coolant mass tracking method and core exit thermocouples (CETs) which are necessary to precise operator actions for accident management (AM) during a loss-of-coolant accident (LOCA) at a pressurized water reactor (PWR). The experiments at the ROSA-V/LSTF facility of the Japan Atomic Energy Research Institute simulated small break LOCAs at the PWR vessel bottom and clarified effects of secondary depressurization as one of the AM measures in case of high pressure injection system failure and non-condensable gas inflow from the accumulator injection system. It was shown that the coolant mass tracking method based on three types of water level instruments could detect most of the primary coolant mass change between the initial state and core-heatup starting condition. The CET characteristics to detect the core heatup conditions were significantly degraded by the condensed water fall-back during the secondary depressurization action.

JAEA Reports

Experimental study on secondary depressurization action for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V/LSTF test SB-PV-03)

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

JAERI-Research 2005-014, 170 Pages, 2005/06

JAERI-Research-2005-014.pdf:7.64MB

A small break LOCA (SBLOCA) experiment was conducted at the LSTF of ROSA-V program to study effects of accident management (AM) on core cooling, which is important in case of high pressure injection (HPI) system failure during an SBLOCA at a PWR. The experiment, SB-PV-03, simulated ten instrument-tube break LOCA at the PWR vessel bottom equivalent to 0.2% cold leg break, total HPI failure, non-condensable gas inflow from accumulator injection system (AIS) and AM actions on secondary depressurization at -55 K/h and auxiliary feedwater (AFW) supply for 30 minutes. It was clarified that the AM actions were effective on primary depressurization until AIS injection end at 1.6 MPa, but thereafter became less effective by the gas inflow, resulting in low pressure injection (LPI) delay and whole core heatup under continuous water discharge at the break. The report describes these phenomena including core heatup related with primary coolant mass and AM actions, primary-to-secondary heat transfer analysis and estimation of gas in the primary loops.

Journal Articles

Thermal-hydraulic responses during PWR pressure vessel upper head small break LOCA based on LSTF experiment and analysis

Takeda, Takeshi; Asaka, Hideaki; Suzuki, Mitsuhiro; Nakamura, Hideo

Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13) (CD-ROM), 8 Pages, 2005/05

no abstracts in English

Journal Articles

Effects of secondary depressurization on PWR bottom small break LOCA experiments in case of HPI failure and non-condensable gas inflow

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 14 Pages, 2004/10

no abstracts in English

Journal Articles

Experimental study on thermal-hydraulics and neutronics coupling effect on flow instability in a heated channel with THYNC facility

Iguchi, Tadashi; Shibamoto, Yasuteru; Asaka, Hideaki; Nakamura, Hideo

Proceedings of 10th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-10), 16 Pages, 2003/10

Thermal-hydraulic and neutronic dynamics are always interrelated in BWR core. This is called thermal-hydraulic and neutronic (T/N) coupling. Channel stability experiments with T/N coupling under non-nuclear condition are very limited. This is mainly due to the difficulties in the real-time simulation of neutron dynamics and in the fast-response void fraction measurement under high-pressure and temperature conditions. Authors have developed techniques to solve the above difficulties, and have succeeded in experimentally simulating T/N coupling under non-nuclear conditions with the THYNC facility. Using THYNC facility, T/N coupling effect on channel stability was investigated. Experiments were performed under Pressure=2-7MPa, Subcooling=10-40K, and Mass flux=270-660kg/m$$^{2}$$s. THYNC results indicated T/N coupling lowered the channel stability threshold. The reduction of channel stability threshold due to T/N coupling was small within 10% at 7MPa in the present THYNC experiment, although the experimental condition was set to be more severe than that supposed in a reactor.

Journal Articles

Study of accident management measures for prevention of severe core damage in ROSA-V program

Asaka, Hideaki; Anoda, Yoshinari

Konsoryu, 17(2), p.116 - 125, 2003/06

no abstracts in English

Journal Articles

Experimental study on cooling limit under flow instability in boiling flow channel

Iguchi, Tadashi; Shibamoto, Yasuteru; Asaka, Hideaki; Nakamura, Hideo

Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 8 Pages, 2003/04

Authors investigated the cooling limit under flow instability, by conducting THYNC experiments using a 2$$times$$2 bundle test section of electrical rod heaters、whose heated lengths and diameters were 3.71m and 12.3mm. The experimental result indicated periodic rise and rapid drop of the rod temperature under flow oscillation, indicating periodic film boiling. When the heating power increased further, the rod temperature indicated continuous film boiling. The power at the onset of continuous film boiling (cooling limit) under flow oscillation was about 50%-80% of the cooling limit under steady flow condition in THYNC. The ratio of both cooling limits almost agreed with the Umekawa model prediction in cases of P$$<$$2MPa and G$$<$$400kg/m2s. For high pressure and high mass flux conditions, the ratio almost agreed with the empirical model based on the heat balance during one cycle of flow oscillation. TRAC-BF1 code simulated periodic film boiling qualitatively, but the cooling limit under the flow oscillation was not predicted well probably due to inaccurate rewetting prediction.

JAEA Reports

Flow resistance of orifices and spacers of BWR thermal-hydraulic and neutronic coupling loop

Iguchi, Tadashi; Asaka, Hideaki; Nakamura, Hideo

JAERI-Research 2002-006, 152 Pages, 2002/03

JAERI-Research-2002-006.pdf:5.52MB

no abstracts in English

Journal Articles

Calculation of thermal-hydraulic behavior of horizontal heat exchanger with parallel tubes

Kondo, Masaya; Otani, Etsuo*; Nakamura, Hideo; Asaka, Hideaki*; Anoda, Yoshinari

Proceedings of 2nd Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-2), p.344 - 350, 2000/00

no abstracts in English

Journal Articles

Single U-tube Testing and RELAP5 code analysis of PCCS with horizontal heat exchanger

Nakamura, Hideo; Kondo, Masaya; Asaka, Hideaki; Anoda, Yoshinari; Tabata, Hiroaki*; Obata, Hiroyuki*

Proceedings of 2nd Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-2), p.336 - 343, 2000/00

no abstracts in English

39 (Records 1-20 displayed on this page)