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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.

Journal Articles

High temperature gas-cooled reactors

Takeda, Tetsuaki*; Inagaki, Yoshiyuki; Aihara, Jun; Aoki, Takeshi; Fujiwara, Yusuke; Fukaya, Yuji; Goto, Minoru; Ho, H. Q.; Iigaki, Kazuhiko; Imai, Yoshiyuki; et al.

High Temperature Gas-Cooled Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.5, 464 Pages, 2021/02

As a general overview of the research and development of a High Temperature Gas-cooled Reactor (HTGR) in JAEA, this book describes the achievements by the High Temperature Engineering Test Reactor (HTTR) on the designs, key component technologies such as fuel, reactor internals, high temperature components, etc., and operational experience such as rise-to-power tests, high temperature operation at 950$$^{circ}$$C, safety demonstration tests, etc. In addition, based on the knowledge of the HTTR, the development of designs and component technologies such as high performance fuel, helium gas turbine and hydrogen production by IS process for commercial HTGRs are described. These results are very useful for the future development of HTGRs. This book is published as one of a series of technical books on fossil fuel and nuclear energy systems by the Power Energy Systems Division of the Japan Society of Mechanical Engineers.

JAEA Reports

Data report of ROSA/LSTF experiment SB-SL-01; Main steam line break accident

Takeda, Takeshi

JAEA-Data/Code 2020-019, 58 Pages, 2021/01

JAEA-Data-Code-2020-019.pdf:3.85MB

An experiment denoted as SB-SL-01 was conducted on March 27, 1990 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-IV (ROSA-IV) Program. The ROSA/LSTF experiment SB-SL-01 simulated a main steam line break (MSLB) accident in a pressurized water reactor (PWR). The test assumptions were made such as auxiliary feedwater (AFW) injection into secondary-side of both steam generators (SGs) and coolant injection from high pressure injection (HPI) system of emergency core cooling system into cold legs in both loops. The MSLB led to a fast depressurization of broken SG, which caused a decrease in the broken SG secondary-side wide-range liquid level. The broken SG secondary-side wide-range liquid level recovered because of the AFW injection into the broken SG secondary-side. The primary pressure temporarily decreased a little just after the MSLB, and increased up to 16.1 MPa following the closure of the SG main steam isolation valves. Coolant was manually injected from the HPI system into cold legs in both loops a few minutes after the primary pressure reduced to below 10 MPa. The primary pressure raised due to the HPI coolant injection, but was kept at less than 16.2 MPa by fully opening a power-operated relief valve of pressurizer. The core was filled with subcooled liquid through the experiment. Thermal stratification was seen in intact loop cold leg during the HPI coolant injection owing to the flow stagnation. On the other hand, significant natural circulation prevailed in broken loop. When the continuous core cooling was ensured by the successive coolant injection from the HPI system, the experiment was terminated. The experimental data obtained would be useful to consider recovery actions and procedures in the multiple fault accident with the MSLB of PWR. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment SB-SL-01.

Journal Articles

Major outcomes through recent ROSA/LSTF experiments and future plans

Takeda, Takeshi; Wada, Yuki; Shibamoto, Yasuteru

World Journal of Nuclear Science and Technology, 11(1), p.17 - 42, 2021/01

Journal Articles

Development of dose estimation system integrating sediment model for recycling radiocesium-contaminated soil to coastal reclamation

Miwa, Kazuji; Takeda, Seiji; Iimoto, Takeshi*

Radiation Protection Dosimetry, 184(3-4), p.372 - 375, 2019/10

 Times Cited Count:0 Percentile:0.01(Environmental Sciences)

The Ministry of the Environment has indicated the policy of recycling the contaminated soil generated by decontamination activity after the Fukushima accident. By recycling to coastal reclamation which is one of effective recycling application, dissolved radiocesium and absorbed radiocesium on soil particles will flow out to the ocean by construction, therefore evaluating radiocesium transition in ocean considering the both types of radiocesium is important for safety assessment. In this study, the radiocesium outflow during constructing and after constructing is modeled, and radiocesium transition in ocean is evaluated by Sediment model suggested in OECD/NEA. The adaptability of sediment model is confirmed by reproducing evaluation of the coastal area of Fukushima. We incorporate the sediment model to PASCLR2 code system to evaluate the doses from radiocesium in ocean.

Journal Articles

Enhancement of element production by incomplete fusion reaction with weakly bound deuteron

Wang, H.*; Otsu, Hideaki*; Chiga, Nobuyuki*; Kawase, Shoichiro*; Takeuchi, Satoshi*; Sumikama, Toshiyuki*; Koyama, Shumpei*; Sakurai, Hiroyoshi*; Watanabe, Yukinobu*; Nakayama, Shinsuke; et al.

Communications Physics (Internet), 2(1), p.78_1 - 78_6, 2019/07

 Times Cited Count:5 Percentile:63.62(Physics, Multidisciplinary)

Searching for effective pathways for the production of proton- and neutron-rich isotopes through an optimal combination of reaction mechanism and energy is one of the main driving forces behind experimental and theoretical nuclear reaction studies as well as for practical applications in nuclear transmutation of radioactive waste. We report on a study on incomplete fusion induced by deuteron, which contains one proton and one neutron with a weak binding energy and is easily broken up. This reaction study was achieved by measuring directly the cross sections for both proton and deuteron for $$^{107}$$Pd at 50 MeV/u via inverse kinematics technique. The results provide direct experimental evidence for the onset of a cross-section enhancement at high energy, indicating the potential of incomplete fusion induced by loosely-bound nuclei for creating proton-rich isotopes and nuclear transmutation of radioactive waste.

Journal Articles

ROSA/LSTF test on nitrogen gas behavior during reflux condensation in PWR and RELAP5 code analyses

Takeda, Takeshi; Otsu, Iwao

Mechanical Engineering Journal (Internet), 5(4), p.18-00077_1 - 18-00077_14, 2018/08

Journal Articles

Uncertainty analysis of ROSA/LSTF test by RELAP5 code and PKL counterpart test concerning PWR hot leg break LOCAs

Takeda, Takeshi; Otsu, Iwao

Nuclear Engineering and Technology, 50(6), p.829 - 841, 2018/08

 Times Cited Count:6 Percentile:73(Nuclear Science & Technology)

JAEA Reports

Data report of ROSA/LSTF experiment SB-SG-10; Recovery actions from multiple steam generator tube rupture accident

Takeda, Takeshi

JAEA-Data/Code 2018-004, 64 Pages, 2018/03

JAEA-Data-Code-2018-004.pdf:3.33MB

Experiment SB-SG-10 was conducted on November 17, 1992 using LSTF. Experiment simulated recovery actions from multiple steam generator (SG) tube rupture accident in PWR. Primary pressure was kept higher than broken SG secondary-side pressure due to coolant injection from high pressure injection (HPI) system into cold and hot legs even after start of full opening of intact SG relief valve (RV). Full opening of power-operated relief valve (PORV) in pressurizer (PZR) resulted in pressure equalization between primary and broken SG systems as well as PZR liquid level recovery. Broken SG RV opened once after start of intact SG RV full opening. Core was filled with saturated or subcooled liquid through experiment. Significant natural circulation prevailed in intact loop after start of intact SG RV full opening. Significant thermal stratification appeared in hot legs especially during time period of HPI coolant injection into hot legs.

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.

Journal Articles

ROSA/LSTF tests and posttest analyses by RELAP5 code for accident management measures during PWR station blackout transient with loss of primary coolant and gas inflow

Takeda, Takeshi; Otsu, Iwao

Science and Technology of Nuclear Installations, 2018, p.7635878_1 - 7635878_19, 2018/00

 Times Cited Count:2 Percentile:33.97(Nuclear Science & Technology)

Journal Articles

RELAP5 uncertainty evaluation using ROSA/LSTF test data on PWR 17% cold leg intermediate-break LOCA with single-failure ECCS

Takeda, Takeshi; Otsu, Iwao

Annals of Nuclear Energy, 109, p.9 - 21, 2017/11

 Times Cited Count:4 Percentile:51.35(Nuclear Science & Technology)

Journal Articles

ROSA/LSTF test and RELAP5 analyses on PWR cold leg small-break LOCA with accident management measure and PKL counterpart test

Takeda, Takeshi; Otsu, Iwao

Nuclear Engineering and Technology, 49(5), p.928 - 940, 2017/08

 Times Cited Count:3 Percentile:41.49(Nuclear Science & Technology)

Journal Articles

ROSA/LSTF test on nitrogen gas behavior during reflux cooling in PWR and RELAP5 post-test analysis

Takeda, Takeshi; Otsu, Iwao

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 11 Pages, 2017/07

JAEA Reports

Synthesized research report in the second mid-term research phase, Mizunami Underground Research Laboratory Project, Horonobe Underground Research Laboratory Project and Geo-stability Project (Translated document)

Hama, Katsuhiro; Sasao, Eiji; Iwatsuki, Teruki; Onoe, Hironori; Sato, Toshinori; Fujita, Tomo; Sasamoto, Hiroshi; Matsuoka, Toshiyuki; Takeda, Masaki; Aoyagi, Kazuhei; et al.

JAEA-Review 2016-014, 274 Pages, 2016/08

JAEA-Review-2016-014.pdf:44.45MB

We synthesized the research results from the Mizunami/Horonobe Underground Research Laboratories (URLs) and geo-stability projects in the second midterm research phase. This report can be used as a technical basis for the Nuclear Waste Management Organization of Japan/Regulator at each decision point from siting to beginning of disposal (Principal Investigation to Detailed Investigation Phase).

JAEA Reports

Data report of ROSA/LSTF experiment TR-LF-07; Loss-of-feedwater transient with primary feed-and-bleed operation

Takeda, Takeshi

JAEA-Data/Code 2016-004, 59 Pages, 2016/07

JAEA-Data-Code-2016-004.pdf:3.34MB

The TR-LF-07 test simulated a loss-of-feedwater transient in a PWR. A SI signal was generated when steam generator (SG) secondary-side collapsed liquid level decreased to 3 m. Primary depressurization was initiated by fully opening a power-operated relief valve (PORV) of pressurizer (PZR) 30 min after the SI signal. High pressure injection (HPI) system was started in loop with PZR 12 s after the SI signal, while it was initiated in loop without PZR when the primary pressure decreased to 10.7 MPa. The primary and SG secondary pressures were kept almost constant because of cycle opening of the PZR PORV and SG relief valves. The PZR liquid level began to drop steeply following the PORV full opening, which caused liquid level formation at the hot leg. The primary pressure became lower than the SG secondary pressure, which resulted in the actuation of accumulator (ACC) system in both loops. The primary feed-and-bleed operation was effective to core cooling because of no core uncovery.

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

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

Takeda, Takeshi; Onuki, Akira*; Kanamori, Daisuke*; Otsu, Iwao

Science and Technology of Nuclear Installations, 2016, p.7481793_1 - 7481793_15, 2016/00

AA2016-0048.pdf:5.15MB

 Times Cited Count:1 Percentile:14.48(Nuclear Science & Technology)

Journal Articles

ROSA/LSTF experiment on a PWR station blackout transient with accident management measures and RELAP5 analyses

Takeda, Takeshi; Otsu, Iwao

Mechanical Engineering Journal (Internet), 2(5), p.15-00132_1 - 15-00132_15, 2015/10

Journal Articles

Thermal hydraulic safety research at JAEA after the Fukushima Dai-ichi Nuclear Power Station accident

Yonomoto, Taisuke; Shibamoto, Yasuteru; Takeda, Takeshi; Satou, Akira; Ishigaki, Masahiro; Abe, Satoshi; Okagaki, Yuria; Sun, Haomin; Tochio, Daisuke

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.5341 - 5352, 2015/08

144 (Records 1-20 displayed on this page)