Data report of ROSA/LSTF experiment SB-PV-03; 0.2% pressure vessel bottom break LOCA with SG depressurization and gas inflow

Takeda, Takeshi

JAEA-Data/Code 2024-014, 76 Pages, 2024/12

JAEA-Data-Code-2024-014.pdf:4.0MB

An experiment denoted as SB-PV-03 was conducted on November 19, 2002 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-PV-03 simulated a 0.2% pressure vessel bottom small-break loss-of-coolant accident in a pressurized water reactor (PWR). The test assumptions included total failure of high pressure injection system of emergency core cooling system (ECCS) and noncondensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of ECCS. Secondary-side depressurization of both steam generators (SGs) as an accident management (AM) action to achieve the depressurization rate of 55 K/h in the primary system was initiated 10 min after the generation of a safety injection signal, and continued afterwards. Auxiliary feedwater injection into the secondary-side of both SGs was started for 30 min with some delay after the onset of the AM action. The AM action was effective on the primary depressurization until the ACC tanks began to discharge nitrogen gas into the primary system. The core liquid level recovered in oscillative manner because of intermittent coolant injection from the ACC system into both cold legs. Therefore, the core liquid level remained at a small drop. The pressure difference between the primary and SG secondary sides became larger after nitrogen gas ingress. Core uncovery occurred by core boil-off during reflux condensation in the SG U-tubes under nitrogen gas influx. When the maximum cladding surface temperature of simulated fuel rods exceeded the pre-determined value of 908 K, the core power was automatically reduced to protect the LSTF core. After the automatic core power reduction, coolant injection from low pressure injection (LPI) system of ECCS into both cold legs led to the whole core quench. After the continuous core cooling was confirmed through the actuation of the LPI system, the experiment was terminated.

Thermal-hydraulic characteristics of coolant in the core bottom structure of the High-Temperature Engineering Test Reactor

Inagaki, Yoshiyuki; Kunitomi, Kazuhiko; ; Ioka, Ikuo*;

Nuclear Technology, 99, p.90 - 103, 1992/07

https://doi.org/10.13182/NT92-A34706

no abstracts in English

Study for thermal mixing of coolant in core bottom structure of HTTR, 2nd Report; Without mixing promoter

Inagaki, Yoshiyuki;

JAERI-M 91-002, 22 Pages, 1991/02

JAERI-M-91-002.pdf:0.74MB

no abstracts in English

Thermal and hydraulic tests of test sections in helium engineering demonstration loop

; Suzuki, Kunihiro; Hino, Ryutaro; Takase, Kazuyuki; Inagaki, Yoshiyuki; Ioka, Ikuo

Nucl. Eng. Des., 120, p.435 - 445, 1990/00

https://doi.org/10.1016/0029-5493(90)90393-C

no abstracts in English

Numerical and experimental study for thermal mixing of coolant in a core-bottom structure of HTTR

Inagaki, Yoshiyuki; Kunugi, Tomoaki;

Nihon Kikai Gakkai Rombunshu, B, 55(514), p.1692 - 1697, 1989/06

no abstracts in English

Performance test of core-bottom insulation structure of VHTR

; ; ;

Nihon Genshiryoku Gakkai-Shi, 27(12), p.1133 - 1135, 1985/00

https://doi.org/10.3327/jaesj.27.1133

no abstracts in English