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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.

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

Large-scaled non-thermal laser peeling, cutting and drilling in nuclear decommissioning industry

Minehara, Eisuke; Hajima, Ryoichi; Sawamura, Masaru; Nagai, Ryoji; Kikuzawa, Nobuhiro; Nishimori, Nobuyuki; Iijima, Hokuto; Nishitani, Tomohiro; Kimura, Hideaki*; Oguri, Daiichiro*; et al.

Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13) (CD-ROM), 10 Pages, 2005/05

The JAERI FEL has recently discovered the new FEL lasing of 255fs ultra fast pulse, 6-9% high-efficiency, one gigawatt high peak power, a few kilowatts average power, and wide tunability of medium and far infrared wavelength regions at the same time. Using the new lasing and energy-recovery linac technology, we could extend a more powerful and more efficient free-electron laser (FEL) than 10kW and 25%, respectively, for nuclear industry, pharmacy, medical, defense, shipbuilding, semiconductor industry, chemical industries, environmental sciences, space-debris, power beaming and so on. In order to realize such a tunable, highly-efficient, high average power, high peak power and ultra-short pulse FEL, we need the efficient and powerful FEL driven by the JAERI compact, stand-alone and zero boil-off super-conducting RF linac with an energy-recovery geometry. Our discussions on the FEL will cover the application of non-thermal peeling, cutting, and drilling to decommission the nuclear power plants, and to prevent stress-corrosion cracking in nuclear industry and roadmap for the industrial FELs, the JAERI compact, stand-alone and zero-boil-off cryostat concept and operational experience, the new, highly-efficient, high-power, and ultra fast pulse lasing mode, and the energy-recovery geometry.

Journal Articles

Highly efficient and high-power industrial FELs driven by a compact, stand-alone and zero-boil-off superconducting rf Linac

Minehara, Eisuke

Nuclear Instruments and Methods in Physics Research A, 483(1-2), p.8 - 13, 2002/05

 Times Cited Count:21 Percentile:78.92(Instruments & Instrumentation)

In order to realize a tunable, highly-efficient, high average power, high peak power and ultra-short pulse free-electron laser(FEL) as a supertool [1]of the 21st for all , the JAERI FEL group and I have developed an industrial FEL driven by a compact, stand-alone and zero-boil-off super-conducting rf linac[2] with an energy-recovery geometry as a conceptual design. Our discussions on the supertool will cover market-requirements for the industrial FELs, some answers from the JAERI compact, stand-alone and zero-boil-off cryostat concept and operational experience over these 8 years, our discovery of the new, highly-efficient, high-power, and ultra-short pulse lasing mode[3], and the energy-recovery geometry.

Journal Articles

Demonstration of the highly-efficient and high-power FELs driven by a superconducting rf linac

Minehara, Eisuke; Hajima, Ryoichi; Sawamura, Masaru; Nagai, Ryoji; Nishimori, Nobuyuki; Kikuzawa, Nobuhiro; Sugimoto, Masayoshi; Yamauchi, Toshihiko; Hayakawa, Takehito; Shizuma, Toshiyuki

Proceedings of 13th Symposium on Accelerator Science and Technology, p.150 - 154, 2001/10

We need a powerful and efficient free-electron laser(FEL) for industrial uses, for examples, pharmacy, medical, defense, shipbuilding, semiconductor industry, chemical industries, environmental sciences, space-debris, power beaming and so on. In order to realize such a tunable, highly-efficient, high average power, high peak power and ultra-short pulse FEL, the JAERI FEL group and I have successfully demonstrated the efficient and powerful FEL driven by a compact, stand-alone and zero-boil-off super-conducting rf linac with an energy-recovery geometry. Our discussions on the FEL will cover market-requirements for the industrial FELs, some answers from the JAERI compact, stand-alone and zero-boil-off cryostat concept and operational experience over these 8 years, our discovery of the new, highly-efficient, high-power, and ultra-short pulse lasing mode, and the energy-recovery geometry.

Journal Articles

A Compact closed loop 2K He3 refrigerator and zero-boil-off cryostat design for the JAERI superconducting RF linac-based FEL

Minehara, Eisuke; Sugimoto, Masayoshi; Sawamura, Masaru; Nagai, Ryoji; Kikuzawa, Nobuhiro; Nishimori, Nobuyuki

Proc. of 11th Symp. on Accelerator Sci. and Technol., P. 236, 1997/00

no abstracts in English

Journal Articles

Measurement of local void fraction distribution in rod bundle under high-pressure high-temperature boil-off conditions by using optical void probe

Kumamaru, Hiroshige; Murata, Hideo; ; Kukita, Yutaka

The 3rd JSME/ASME Joint Int. Conf. on Nuclear Engineering (ICONE),Vol. 1, 0, p.217 - 222, 1995/00

no abstracts in English

Journal Articles

Void-fraction distribution under high-pressure boil-off conditions in rod bundle geometry

Kumamaru, Hiroshige; ; Murata, Hideo; Kukita, Yutaka

Nucl. Eng. Des., 150, p.95 - 105, 1994/00

 Times Cited Count:27 Percentile:89.43(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Critical heat flux for rod bundle under high-pressure boil-off conditions

Guo, Z.*; Kumamaru, Hiroshige; Kukita, Yutaka

JAERI-M 93-238, 20 Pages, 1993/12

JAERI-M-93-238.pdf:0.67MB

no abstracts in English

JAEA Reports

Core void fraction distribution under high-temperature high-pressure boil-off conditions; Experimental study with two-phase flow test facility(TPTF)

; Kumamaru, Hiroshige; Murata, Hideo; Anoda, Yoshinari; Kukita, Yutaka

JAERI-M 93-200, 56 Pages, 1993/10

JAERI-M-93-200.pdf:1.85MB

no abstracts in English

JAEA Reports

Boil-off calorimetric measurement of inductive heating energy in stability test of superconducting magnet

Tsukamoto, Hideo*; *; Koizumi, Norikiyo; Isono, Takaaki; Takahashi, Yoshikazu; Nishi, Masataka

JAERI-M 93-001, 25 Pages, 1993/02

JAERI-M-93-001.pdf:0.64MB

no abstracts in English

Journal Articles

Critical heat flux and heat transfer above mixture level under high-pressure boil-off conditions in PWR type and tight-lattice type fuel bundles

Kumamaru, Hiroshige; Kukita, Yutaka

Nucl. Eng. Des., 144, p.257 - 268, 1993/00

 Times Cited Count:1 Percentile:19.22(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Critical heat flux and heat transfer above mixture level under high-pressure boil-off conditions for PWR type and tight-lattice type fuel bundles

Kumamaru, Hiroshige; Kukita, Yutaka

ANP 92: Proc. of the Int. Conf. on Design and Safety of Advanced Nuclear Power Plants,Vol. 3, p.24.4-1 - 24.4-7, 1992/00

no abstracts in English

Journal Articles

Small break LOCA counter part test in the LSTF, BETHSY, LOBI and SPES test facilities

A.Annunziato*; C.Addabbo*; G.Briday*; R.Deruaz*; D.Juhel*; Kumamaru, Hiroshige; Kukita, Yutaka; C.Medich*; M.Rigamonti*

Proc. of the 5th Int. Topical Meeting on Reactor Thermal Hydraulics: NURETH-5, p.1570 - 1576, 1992/00

no abstracts in English

Journal Articles

Loss of residual heat removal(RHR) during PWR MID-loop operation; Experiments in ROSA-IV/LSTF

Nakamura, Hideo; Anoda, Yoshinari; Kukita, Yutaka

Proc. of the Int. Topical Meeting on Safety of Thermal Reactors, p.497 - 503, 1991/00

no abstracts in English

17 (Records 1-17 displayed on this page)
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