Input data preparation for PWR large-break LOCA analysis with RELAP5/MOD3.3 code

Takeda, Takeshi

JAEA-Data/Code 2025-005, 106 Pages, 2025/06

JAEA-Data-Code-2025-005.pdf:2.93MB

JAEA has been creating input data for pressurized water reactor (PWR) analysis with RELAP5/MOD3.3 code, mainly based on design information for the four-loop PWR's Tsuruga Power Station Unit-2 as the reference reactor of the Large Scale Test Facility (LSTF). The cold leg large-break loss-of-coolant accident (LBLOCA) calculation in the flamework of the BEMUSE program is cited as a representative OECD/NEA activity related to the PWR analysis. The new regulatory requirements for PWRs in Japan include the event of loss of recirculation functions from emergency core cooling system (ECCS) in the cold leg LBLOCA. This event should be evaluated the effectiveness of measures against severe core damage. The input data for this study were made preparations to analyze the PWR LBLOCA, which is one of the design basis accidents that should be postulated in the safety design. This report describes the main features of the input data for the PWR LBLOCA analysis. The PWR model comprised a reactor vessel, pressurizer (PZR), hot legs, steam generators (SGs), SG secondary-side system, crossover legs, cold legs, and ECCS. A four-loop PWR was simulated by two loops in the LBLOCA calculation. Specifically, loop-A attached with the PZR corresponded to three loops, and loop-B mounted with the break was equal to one loop. The nodalization schemes of the PWR components were referred to those of the LSTF components. Moreover, interpretations were added to the main input data for the PWR LBLOCA analysis, and further information such as the basis for determining the input data was provided. In addition, transient analysis was performed employing the prepared input data for the loss of ECCS recirculation functions event. The present transient analysis was confirmed to be appropriate generally by comparing with the calculation in the previous study using the RELAP5/MOD3.3 code. Furthermore, sensitivity analyses were executed exploiting the RELAP5/MOD3.3 code to clarify the effects of a discharge coefficient through the break and water injection flow rate of the alternative recirculation on the fuel rod cladding surface temperature. This report explains the results of the sensitivity analyses within the defined ranges, which complement some of the content of the previous study's calculation for the loss of ECCS recirculation functions event.

Effects of high temperature ECC injection on small and large break BWR LOCA simulation tests in ROSA-III program; RUNs 940 and 941

Suzuki, Mitsuhiro; Nakamura, Hideo; Kumamaru, Hiroshige; Anoda, Yoshinari; Yonomoto, Taisuke; Murata, Hideo; Tasaka, Kanji

JAERI-M 90-051, 256 Pages, 1990/03

JAERI-M-90-051.pdf:6.41MB

no abstracts in English

Similarity study of ROSA-III and fist large bleak counterpart tests to BWR large bleak LOCA

; ; ; Tasaka, Kanji; J.A.Findlay*; W.A.Sutherland*

Nucl.Eng.Des., 103, p.223 - 238, 1987/00

https://doi.org/10.1016/0029-5493(87)90276-7

no abstracts in English

Recirculation Pump Suction Line 200% Break Integral Test at ROSA-III with Two LPCI Failures,RUN 983

; ; Tasaka, Kanji; ; ; ; ;

JAERI-M 84-135, 206 Pages, 1984/08

JAERI-M-84-135.pdf:4.94MB

no abstracts in English

Investigation of Similarity Between ROSA-III and BWR/6 During Large Break LOCA

; ; ; Tasaka, Kanji; ;

JAERI-M 83-046, 144 Pages, 1983/03

JAERI-M-83-046.pdf:3.56MB

no abstracts in English

The JAERI code system for evaluation of BWR ECCS performance

; ; Asahi, Yoshiro; ; ; ;

JAERI 1283, 238 Pages, 1982/12

JAERI-1283.pdf:10.48MB

no abstracts in English

ALARM-B2:A Computer Program for Analysis of Large Break LOCA of BWR

JAERI-M 9655, 98 Pages, 1981/09

JAERI-M-9655.pdf:2.16MB

no abstracts in English

ALARM-B1: A Computer Program for Boiling Water Reactor Blowdown Analysis

JAERI-M 6968, 41 Pages, 1977/03

JAERI-M-6968.pdf:1.19MB

no abstracts in English