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.

Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR) (FY2018)

Department of HTTR

JAEA-Review 2019-049, 97 Pages, 2020/03

JAEA-Review-2019-049.pdf:4.66MB

The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor being able to get 950C temperature of the outlet coolant with 30 MW of thermal power, constructed at the Oarai Research and Development Institute of the Japan Atomic Energy Agency is the first High- Temperature Gas-cooled Reactor (HTGR) in Japan. The purpose of the HTTR is to establish and upgrade basic technologies for HTGRs. The HTTR has accumulated a lot of experience of HTGRs' operation and maintenance up to the present time throughout rated power operations, safety demonstration tests, long-term high temperature operations and demonstration tests relevant to HTGRs' R&Ds. In the fiscal year 2018, we made effort to pass the inspection of application document for the HTTR licensing to prove conformity with the new regulatory requirements for research reactors that took effect since December 2013 in order to restart operations of the HTTR that stopped since the 2011 off the Pacific coast of Tohoku Earthquake. This report summarizes the activities carried out in the 2018 fiscal year, which were the situation of the new regulatory requirements screening of the HTTR, the operation and maintenance of the HTTR, R&Ds relevant to commercial-scale HTGRs, the international cooperation on HTGRs and so on.

Safety requirements expected for the prototype fast breeder reactor "Monju"

Saito, Shinzo; Okamoto, Koji*; Kataoka, Isao*; Sugiyama, Kenichiro*; Muramatsu, Ken*; Ichimiya, Masakazu*; Kondo, Satoru; Yonomoto, Taisuke

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 10 Pages, 2015/05

OECD/NEA/CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Code Requirements

Kukita, Yutaka*; Arai, Kenji*;

Nihon Genshiryoku Gakkai-Shi, 39(2), p.151 - 153, 1997/00

no abstracts in English

Proceedings of the 1st JAERI Symposium on HTGR Technologies; Design, Licensing Requirements and Supporting Technologies

JAERI-M 90-103, 399 Pages, 1990/07

JAERI-M-90-103.pdf:12.1MB

no abstracts in English

Current status of JMTR

Kaminaga, Masanori; Araki, Masanori

no journal, , 

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.