Design and modification of steam generator safety system of FBR MONJU

Matsuura, Makoto; Hatori, Masakazu; Ikeda, Makinori

Nuclear Engineering and Design, 237(12-13), p.1419 - 1428, 2007/07

https://doi.org/10.1016/j.nucengdes.2006.09.046

MONJU is a prototype Fast Breeder Reactor (FBR) in Japan. The sodium-water reaction in the steam generator (SG) is one of the important safety assessment items for a sodium cooled reactor like MONJU. A latest evaluation method on the design margin against the overheating tube rupture showed that it had not enough margins on the present SG system in the worst case. For improving that margin, it needs to shorten the time of the sodium-water reaction by earlier water leak detection in the SG and sooner water ejection from the SG tubes. Therefore MONJU is now carrying out the following modification works (1) addition of steam relief valves, (2) addition of a gas pressure gauge with changing the interlock logic and lowering the trigger level, (3) reducing the opening of the valves on the SG gas flow line to the dump tank because of earlier detection for the pressure rise. After this modification, the design margin of the SG system will be sufficiently improved. For the tube materials, the value of the cumulative damage for the creep failure will be much smaller than 1, and the equivalent stress for the ductile failure will be much smaller than the tensile strength.

Evaluation of mechanical integrity of floor liner in secondary heat transport system cells of Monju

Imo, Kazumichi; Onuki, Koji; Kikuchi, H.; Morishita, Masaki; Igawa, Kenichi*; Nishibayashi, Yohei; Ikeda, Makinori

JNC TN2400 2003-004, 78 Pages, 2004/03

JNC-TN2400-2003-004.pdf:4.74MB

About the secondary system floor liner of Monju, the mechanical soundness of a floor liner to the thermal load conditions by combustion of leak sodium was evaluated by large displacement inelastic analysis and the partial structure model test.In large displacement inelastic analysis, it was confirmed that heat strain of a floor liner was less than the standard value of the strain for no through-wall crack. In addition, the influence evaluation by a temperature increasing rate, liner board thickness, and the existence of corrosion thinning was performed. The result, the influence on the maximum strain was small in any parameter.In the partial structure model test, even when a strain more excessive than the standard value of strain was given, it was confirmed no through-wall crack in a liner plate.In addition, this report improves the evaluation conditions of the conventional research report [1], and resummarizes them to detailed design evaluation.

Design and characteristics of annulus ventilation and HVAC system for prototype FBR MONJU

Ikeda, Makinori

Proceedings of 22nd DOE/NRC Nuclear Air Cleaning Conference, p.791 - 807, 1993/07

None

None

*; *; Otani, Nobuo*; ; *; *; *

PNC TN9410 89-144, 132 Pages, 1989/10

PNC-TN9410-89-144.pdf:3.47MB

None

Confirmation plan of the integrity of the plant facilities after a long time shutdown at prototype fast breeder reactor MONJU

Ito, Kazumoto; Maeda, Hiroshi; Ikeda, Makinori

no journal, , 

no abstracts in English

Evaluation of seismic safety of the prototype fast breeder reactor Monju on the basis of the revised regulatory guide for seismic design, 1; Assessment of the design basis ground motions and overview of re-evaluation of seismic safety related facilities

Ohara, Takaharu; Ikeda, Makinori; Morishita, Masaki; Miyazaki, Masashi; Shimada, Koji; Moriizumi, Makoto*

no journal, , 

no abstracts in English

Current state of the additional geological surveys of crush zones at the fast breeder prototype reactor "Monju" site

Ishimaru, Tsuneari; Shimada, Koji; Sasaki, Akimichi; Tanaka, Yukumo; Miyazaki, Masashi; Yasue, Kenichi; Niwa, Masakazu; Sueoka, Shigeru; Umeda, Koji; Ikeda, Makinori

no journal, , 

The Nuclear Regulation Authority instructed to develop a further additional research plan regarding the crushed zones on the Monju site against the JAEA on September 25. The JAEA developed a research plan in response to this instruction, and, the stripping investigation with expansion of the area, the additional detailed geographical and geological surveys around the mountains/terrace boundary, and the marine seismic surveys in the coastal waters, were conducted. The fracture zones were grouped into 2 systems called -system and -system in the on-site investigation. We examined the cross-cutting relationships and displacements of the fracture zones and confirmed that the -system was formed after the -system. From the investigation results so far, similar to the results of the compilation report of April 31, 2010, no clear evidence was observed to support that on-site crushed zones are active faults.

Evaluation of double leakage at primary heat transport systems of Monju with passive safety features

Yoshimura, Kazuo; Ikeda, Makinori; Enuma, Yasuhiro; Aizawa, Kosuke

no journal, , 

JAEA has conducted a safety evaluation of impacts of double leakage at the PHTSs considering the passive safety features of Monju as the best-estimate evaluation for a DEC. The result shows that the total amount of leaked sodium can be reduced by the depressurization of the cover gas resulting from decrease in coolant inventory, i.e. negative pressure effects. The reactor coolant level required for decay heat removal, therefore, can be maintained even under double leakage at the PHTSs.