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JAEA Reports

Data report of ROSA/LSTF experiment SB-PV-09; 1.9% pressure vessel top small break LOCA with SG depressurization and gas inflow

Takeda, Takeshi

JAEA-Data/Code 2021-006, 61 Pages, 2021/04

JAEA-Data-Code-2021-006.pdf:2.78MB

An experiment denoted as SB-PV-09 was conducted on November 17, 2005 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-PV-09 simulated a 1.9% pressure vessel top small-break loss-of-coolant accident in a pressurized water reactor (PWR). The test assumptions included total failure of high pressure injection system and non-condensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of emergency core cooling system (ECCS). In the experiment, liquid level in the upper-head was found to control break flow rate. When maximum core exit temperature reached 623 K, steam generator (SG) secondary-side depressurization was initiated by fully opening the relief valves in both SGs as an accident management (AM) action. The AM action, however, was ineffective on the primary depressurization until the SG secondary-side pressure decreased to the primary pressure. Meanwhile, the core power was automatically reduced when maximum cladding surface temperature of simulated fuel rods exceeded the pre-determined value of 958 K to protect the LSTF core due to late and slow response of core exit temperature. After the automatic core power reduction, loop seal clearing (LSC) was induced in both loops by steam condensation on the ACC coolant injected into cold legs. The whole core was quenched because of core recovery after the LSC. After the ACC tanks started to discharge nitrogen gas, the pressure difference between the primary and SG secondary sides became larger. After the continuous core cooling was confirmed through the actuation of low pressure injection system of ECCS, the experiment was terminated. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment SB-PV-09.

JAEA Reports

Data report of ROSA/LSTF experiment SB-PV-07; 1% Pressure vessel top break LOCA with accident management actions and gas inflow

Takeda, Takeshi

JAEA-Data/Code 2018-003, 60 Pages, 2018/03

JAEA-Data-Code-2018-003.pdf:3.68MB

Experiment SB-PV-07 was conducted on June 9, 2005 using LSTF. Experiment simulated 1% pressure vessel top small-break LOCA in PWR under total failure of HPI system and nitrogen gas inflow to primary system from ACC tanks. Liquid level in upper-head was found to control break flow rate. Coolant was started to manually inject from HPI system into cold legs as first accident management (AM) action when maximum core exit temperature reached 623 K. Fuel rod surface temperature largely increased because of late and slow response of core exit temperature. SG secondary-side depressurization was initiated by fully opening relief valves as second AM action when primary pressure decreased to 4 MPa. However, second AM action was not effective on primary depressurization until SG secondary-side pressure decreased to primary pressure. Pressure difference became larger between primary and SG secondary sides after ACC tanks started to discharge nitrogen gas.

JAEA Reports

Data report of ROSA/LSTF experiment SB-HL-12; 1% Hot leg break LOCA with SG depressurization and gas inflow

Takeda, Takeshi

JAEA-Data/Code 2015-022, 58 Pages, 2016/01

JAEA-Data-Code-2015-022.pdf:3.31MB

The SB-HL-12 test simulated PWR 1% hot leg SBLOCA under assumptions of total failure of HPI system and non-condensable gas (nitrogen gas) inflow. SG depressurization by fully opening relief valves in both SGs as AM action was initiated immediately after maximum fuel rod surface temperature reached 600 K. After AM action due to first core uncovery by core boil-off, the primary pressure decreased, causing core mixture level swell. The fuel rod surface temperature then increased up to 635 K. Second core uncovery by core boil-off took place before LSC induced by steam condensation on ACC coolant injected into cold legs. The core liquid level recovered rapidly after LSC. The fuel rod surface temperature then increased up to 696 K. The pressure difference became larger between the primary and SG secondary sides after nitrogen gas inflow. Third core uncovery by core boil-off occurred during reflux condensation. The maximum fuel rod surface temperature exceeded 908 K.

Journal Articles

RELAP5 code study of ROSA/LSTF experiments on PWR safety system using steam generator secondary-side depressurization

Takeda, Takeshi; Onuki, Akira*; Nishi, Hiroaki*

Journal of Energy and Power Engineering, 9(5), p.426 - 442, 2015/05

Journal Articles

RELAP5 code study of ROSA/LSTF validation tests for PWR safety system using SG secondary-side depressurization

Takeda, Takeshi; Onuki, Akira*; Nishi, Hiroaki*

Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 13 Pages, 2014/12

JAEA Reports

Data report of ROSA/LSTF experiment SB-CL-32; 1% cold leg break LOCA with SG depressurization and no gas inflow

Takeda, Takeshi

JAEA-Data/Code 2014-021, 59 Pages, 2014/11

JAEA-Data-Code-2014-021.pdf:5.16MB

Experiment SB-CL-32 was conducted on May 28, 1996 using the LSTF. The experiment SB-CL-32 simulated 1% cold leg small-break LOCA in PWR under assumptions of total failure of HPI system and no inflow of non-condensable gas from ACC tanks. Secondary-side depressurization of both SGs as AM action to achieve the depressurization rate of 200 K/h in the primary system was initiated 10 min after break. Core uncovery started with liquid level drop in crossover leg downflow-side. The core liquid level recovered rapidly after first LSC. The surface temperature of simulated fuel rod then increased up to 669 K. Core uncovery took place before second LSC induced by steam condensation on ACC coolant. The core liquid level recovered rapidly after second LSC. The maximum fuel rod surface temperature was 772 K. The continuous core cooling was confirmed because of coolant injection by LPI system. This report summarizes the test procedures, conditions and major observation.

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:59.81(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

Nihon 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.

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

JAEA Reports

Journal Articles

Secondary-side depressurization during PWR cold-leg small break LOCAs based on ROSA-V/LSTF experiments and analyses

Asaka, Hideaki; Anoda, Yoshinari; Kukita, Yutaka*;

Journal of Nuclear Science and Technology, 35(12), p.905 - 915, 1998/12

 Times Cited Count:15 Percentile:76.16(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Core makeup tank behavior observed during the ROSA-AP600 experiments

Yonomoto, Taisuke; ; Kukita, Yutaka; L.S.Ghan*; R.R.Schultz*

Nuclear Technology, 119, p.112 - 122, 1997/08

 Times Cited Count:23 Percentile:85.12(Nuclear Science & Technology)

no abstracts in English

Journal Articles

PWR small break Loss-of-Coolant-Accident Experiment at ROSA-V/LSTF with a combination of secondary-side depressurization and gravity-driven safety injection

Yonomoto, Taisuke; ; Kukita, Yutaka

Journal of Nuclear Science and Technology, 34(6), p.571 - 581, 1997/06

 Times Cited Count:5 Percentile:43.99(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Small break LOCA tests at ROSA-V/LSTF on next generation PWR designs

Yonomoto, Taisuke; ; ; Anoda, Yoshinari; Kukita, Yutaka*

Eighth Int. Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), 1, p.535 - 542, 1997/00

no abstracts in English

JAEA Reports

A Study of entrainment at a break and in the core during small break loss-of-coolant accidents in PWRs

Yonomoto, Taisuke

JAERI-Research 96-024, 154 Pages, 1996/05

JAERI-Research-96-024.pdf:5.87MB

no abstracts in English

Journal Articles

Intentional depressurization of steam generator secondary side during a PWR small-break loss-of-coolant accident

; Kukita, Yutaka

Journal of Nuclear Science and Technology, 32(2), p.101 - 110, 1995/02

 Times Cited Count:16 Percentile:81.78(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Integral testing on thermal-hydroulic responces of advanced passive-safety PWR AP600

Anoda, Yoshinari; Kukita, Yutaka; Yonomoto, Taisuke; ; Nakamura, Hideo; Kumamaru, Hiroshige; T.J.Boucher*; M.G.Ortiz*; R.A.Shaw*; R.R.Schultz*

Nihon Kikai Gakkai Dai-72-Ki Tsujo Sokai Koenkai Koen Rombunshu, III, 0, p.413 - 414, 1995/00

no abstracts in English

Journal Articles

Long-term oscillations of pressurizer liquid level observed during ROSA/AP600 experiments

Yonomoto, Taisuke; Kukita, Yutaka

Proc. of the 1995 Int. Joint Power Generation Conf. (NE-Vol. 17), 2, p.25 - 31, 1995/00

no abstracts in English

Journal Articles

RELAP5/MOD3 code analyses of LSTF tests on intentional primary system depressurization following PWR small-break LOCA

Kumamaru, Hiroshige; ; M.Wang*; Kukita, Yutaka

Validation of Systems Transients Analysis Codes (FED-Vol. 223), 0, p.129 - 136, 1995/00

no abstracts in English

79 (Records 1-20 displayed on this page)