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.

Design of high temperature irradiation materials inspection cells (spent fuel inspection cells) in the High Temperature engineering Test Reactor

Ino, Hiroichi*; Ueta, Shohei; Suzuki, Hiroshi; Tobita, Tsutomu*; Sawa, Kazuhiro

JAERI-Tech 2001-083, 46 Pages, 2002/01

JAERI-Tech-2001-083.pdf:6.31MB

no abstracts in English

Current status toward the reoperation of JMTR

Kaminaga, Masanori; Kusunoki, Tsuyoshi; Araki, Masanori

no journal, , 

The Japan Materials Testing Reactor (JMTR) is a light water cooled and moderated tank type research reactor with 50MW thermal power. From its first criticality in March 1968, the JMTR has been utilized for the fuel/material irradiation examinations of the LWRs, the HTGR and nuclear fusion research as well as for the RI productions. The JMTR operation was once stopped in order to have a check & review in August 2006, and the refurbishment and restart of the JMTR was finally determined after the national discussion. The refurbishment was started from JFY 2007, and was finished in March 2011. However, at the end of the JFY 2010, the Great East Japan Earthquake occurred, and the functional tests before the JMTR restart were delayed. On the other hand, based on the safety assessments considering the Great East Japan Earthquake in 2011, new regulatory requirements for research and test reactors have been established on Dec.18, 2013 by the NRA (Nuclear Regulation Authority). The new regulatory requirements include the satisfaction of integrities for the updated earthquake forces, Tsunami, the consideration of natural phenomena, the provision of manuals for full evacuation, and the management of consideration in the Beyond Design Basis Accidents to protect fuel damage and to mitigate impact of the accidents. Above analyses have intensively been performed, and an application to the NRA was submitted on March 27, 2015. In this presentation, the recent activities for the new regulatory requirements and a training course for foreign young researchers and engineers are introduced.