Phase 1 code assessment of SIMMER-III; A Computer program for LMFR core disruptive accident analysis

Kondo, Satoru; Tobita, Yoshiharu

JAEA-Research 2019-009, 382 Pages, 2020/03

JAEA-Research-2019-009.pdf:22.82MB

The SIMMER-III computer code, developed at the Japan Atomic Energy Agency (JAEA, the former Power Reactor and Nuclear Fuel Development Corporation), is a two-dimensional, multi-velocity-field, multi-component fluid-dynamics code, coupled with a space- and time-dependent neutron kinetics model. The code is being used widely for simulating complex phenomena during core-disruptive accidents (CDAs) in liquid-metal fast reactors (LMFRs). In parallel to the code development, a comprehensive assessment program was performed in two phases: Phase 1 for verifying individual fluid-dynamics models; and Phase 2 for validating its applicability to integral phenomena important to evaluating LMFR CDAs. The SIMMERIII assessment program was participated by European research and development organizations, and the achievement of Phase 1 was compiled and synthesized in 1996. This report has been edited by revising and reproducing the original 1996 informal report, which compiled the achievement of Phase 1 assessment. A total of 34 test problems were studied in the areas: fluid convection, interfacial area and momentum exchange, heat transfer, melting and freezing, and vaporization and condensation. The problems identified have been reflected to the Phase 2 assessment and later model development and improvement. Although the revisions were made in the light of knowledge base obtained later, the original individual contributions by the participants, both positive and negative, are retained except for editorial changes.

Radiation monitoring using manned helicopter around the Nuclear Power Station in the fiscal year 2017 (Contract research)

Futemma, Akira; Sanada, Yukihisa; Ishizaki, Azusa; Iwai, Takeyuki*; Seguchi, Eisaku; Matsunaga, Yuki*; Kawabata, Tomoki; Toyoda, Masayuki*; Tobita, Shinichiro*; Hiraga, Shogo*; et al.

JAEA-Technology 2018-015, 120 Pages, 2019/02

JAEA-Technology-2018-015.pdf:15.01MB

By the nuclear disaster of Fukushima Daiichi Nuclear Power Station (FDNPS), Tokyo Electric Power Company (TEPCO), caused by the Great East Japan Earthquake and the following tsunami on March 11, 2011, a large amount of radioactive material was released from the NPS. After the nuclear disaster, airborne radiation monitoring using manned helicopter was conducted around FDNPS. The results in the fiscal 2017 were summarized in this report. In addition, we developed and systemized the discrimination technique of the Rn-progenies. The accuracy of aerial radiation monitoring was evaluated by taking into consideration GPS data error.

Improvement of a physical model for blockage formation of solid-liquid mixture flow with freezing for core safety evaluation of SFRs

Aoyagi, Mitsuhiro; Kamiyama, Kenji; Tobita, Yoshiharu

Journal of Nuclear Science and Technology, 55(5), p.530 - 538, 2018/05

https://doi.org/10.1080/00223131.2017.1417172

Background radiation monitoring using manned helicopter for establishment of technique of nuclear emergency response in the fiscal year 2016 (Contract research)

Sanada, Yukihisa; Mori, Airi; Iwai, Takeyuki; Seguchi, Eisaku; Matsunaga, Yuki*; Kawabata, Tomoki; Toyoda, Masayuki*; Tobita, Shinichiro*; Hiraga, Shogo; Sato, Yoshiharu; et al.

JAEA-Technology 2017-035, 69 Pages, 2018/02

JAEA-Technology-2017-035.pdf:32.92MB

By the nuclear disaster of Fukushima Daiichi Nuclear Power Station (FDNPS), Tokyo Electric Power Company (TEPCO), caused by the East Japan earthquake and the following tsunami occurred on March 11, 2011, a large amount of radioactive materials was released from the NPS. After the nuclear disaster, airborne radiation monitoring using manned helicopter was conducted around FDNPS. We carried out the background monitoring around the nuclear power stations of the whole country to apply a technique of the airborne radiation monitoring that is cultivated in Fukushima as a technology of nuclear emergency response. This result of the aerial radiation monitoring using the manned helicopter around Ooi, Takahama and Ikata Nuclear Power Station and in the fiscal 2016 were summarized in the report. In addition, technical issues were described.

Experimental study on debris bed characteristics for the sedimentation behavior of solid particles used as simulant debris

Shamsuzzaman, M.*; Horie, Tatsuro*; Fuke, Fusata*; Kamiyama, Motoki*; Morioka, Toru*; Matsumoto, Tatsuya*; Morita, Koji*; Tagami, Hirotaka; Suzuki, Toru*; Tobita, Yoshiharu

Annals of Nuclear Energy, 111, p.474 - 486, 2018/01

https://doi.org/10.1016/j.anucene.2017.09.011

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 2; Assessment of PAMR/PAHR phase in ULOF

Sogabe, Joji; Suzuki, Toru; Wada, Yusaku; Tobita, Yoshiharu

Nihon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00393_1 - 16-00393_10, 2017/04

https://doi.org/10.1299/transjsme.16-00393

The achievement of In-Vessel Retention (IVR) of the accident consequences in an unprotected loss of flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, is effective and rational approach in enhancing the safety characteristics of sodium-cooled fast reactor. The objective of the present study is to show that the decay heat generated from the relocated fuels would be stably removed in post-accident-material-relocation/post-accident-heat-removal (PAMR/PAHR) phase, where the relocated fuels mean fuel discharged from the core into the low-pressure plenum through control-rod guide tubes, and fuel remnant in the disrupted core region (non-discharged fuel). As a result of the present assessments, it should be concluded that the stable cooling of the relocated fuels was confirmed and the prospect of IVR was obtained.

Event sequence analysis of core disruptive accident in a metal-fueled sodium-cooled fast reactor

Yamano, Hidemasa; Tobita, Yoshiharu; Kubo, Shigenobu; Ueda, Nobuyuki*

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 10 Pages, 2016/11

In this study, the event sequence analysis of CDA in a large metal-fueled SFR has been performed in order to investigate reactivity progression and molten fuel relocation behavior in the metal-fueled SFR. The initiating phase analysis during the CDA initiated by unprotected loss-of-flow accidents has been conducted using the CANIS code, which showed a small power peak. Using the initial conditions based on the initiating phase analysis, the SIMMER-III code was applied to a whole-core scale analyses to clarify the event sequence including the reactivity progression and molten fuel relocation. As a result, recriticality in the whole core analysis resulted in a very mild energy release. The mild energy release in the metal-fueled core can be attributed to the small specific heat of metal fuel and the large prompt negative reactivity feedback mechanism.

Improvements to the simmer code model for steel wall failure based on EAGLE-1 test results

Toyooka, Junichi; Kamiyama, Kenji; Tobita, Yoshiharu; Suzuki, Toru

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11

An Empirical correlation to predict the distance for fragmentation of simulated Molten-Core materials discharged into a sodium pool

Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 8 Pages, 2016/10

In order to evaluate the distance for fragmentation of molten core material discharged into the lower sodium plenum during core disruptive accidents in sodium-cooled fast reactors, experiments with simulated molten materials and coolants (water, sodium) was carried out, where an empirical correlation of the distance for fragmentation was developed. The empirical correlation developed by this study showed a good agreement with the measurement results obtained by the present experiments. It was found that in order to well-predict the distance for fragmentation in sodium, thermal phenomena, such as sodium boiling and resultant vapor expansion, needed to be considered.

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 1; Overview of IVR evaluation in Anticipated Transient without Scram (ATWS)

Suzuki, Toru; Sogabe, Joji; Tobita, Yoshiharu; Sakai, Takaaki*; Nakai, Ryodai

Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2016/06