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.

Investigations of accelerator reliability and decay heat removal for accelerator-driven system

Sugawara, Takanori; Takei, Hayanori; Tsujimoto, Kazufumi

Annals of Nuclear Energy, 125, p.242 - 248, 2019/03

https://doi.org/10.1016/j.anucene.2018.11.001

To realize the feasible accelerator-driven system (ADS) concept, the investigations for the reliable accelerator and conceptual plant design considering safety issues were performed. As the reliable accelerator concept, the double-accelerator concept was proposed to reduce the beam-trip frequency. The estimated beam-trip frequency with the double-accelerator concept using the J-PARC LINAC operation data showed that the beam-trip frequency was significantly improved with the comparison of the single accelerator result. The basic investigation of the primary reactor auxiliary coolant system (PRACS) was performed for the safety design of the LBE cooled ADS. The concept which the PRACS heat exchanger was integrated to the steam generator was proposed and the transient analysis in the loss of heat sink accident was carried out. The result presented that the decay heat removal was appropriate when the operation of the PRACS succeeded.

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

https://doi.org/10.1299/mej.18-00077

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

https://doi.org/10.1155/2018/7635878

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

https://doi.org/10.1016/j.net.2017.03.004

Applicability confirmation test of optimum decay heat evaluation method for HTGR with HTTR (Non-nuclear heating test); Validation of residual heat evaluation model

Honda, Yuki; Inaba, Yoshitomo; Nakagawa, Shigeaki; Yamazaki, Kazunori; Kobayashi, Shoichi; Aono, Tetsuya; Shibata, Taiju; Ishitsuka, Etsuo

JAEA-Technology 2017-013, 20 Pages, 2017/06

JAEA-Technology-2017-013.pdf:2.52MB

Decay heat is one of an important factor for a safety evaluation of depressurized loss-of-forced cooling accident, a representative high consequence accident, in high temperature gas-cooled reactor (HTGR). Traditionally, a conservative decay heat curve is used for safety analysis according to the regulatory standards. On the other hand, there is growing interest in obtaining test data related to decay heat for the use of uncertainty analysis. However, such data has not been obtained for prismatic-type HTGR. Therefore, we have launched a test program to obtain the decay heat data from the HTTR. As an initial step, an applicability confirmation test of decay heat evaluation method for HTGR was conducted in February 2017 without non-nuclear heating condition. This report introduces an estimation method for the decay heat based on test data using HTTR and shows the results of validation of the reactor residual heat evaluation method which will be used to obtain the decay heat data based on test data.

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

http://doi.org/10.1299/mej.15-00132

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

https://doi.org/10.17265/1934-8975/2015.05.002

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

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

https://doi.org/10.3327/jnst.43.55

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.