Development of failure mitigation technologies for improving resilience of nuclear structures, 6; Resilience improvements of fast reactors by failure mitigation for excessive earthquake

Yamano, Hidemasa; Futagami, Satoshi; Ando, Masanori; Kurisaka, Kenichi

Transactions of 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 11 Pages, 2024/03

In this study, the dynamic structural analysis of the reactor vessel for excessive earthquake using the FINAS/STAR code has shown the elephant foot buckling deformation and calculated the cumulative fatigue failure fraction. Using the calculation results, this paper describes the fragility curve using the safety factor method, indicating the significantly improved curve compared the previous one.

Structural analysis of a reactor vessel in a sodium-cooled fast reactor under extremely high temperature conditions

Yamano, Hidemasa; Futagami, Satoshi; Ando, Masanori

Mechanical Engineering Journal (Internet), 10(4), p.23-00043_1 - 23-00043_12, 2023/08

https://doi.org/10.1299/mej.23-00043

This study has conducted a detailed structural analysis of a reactor vessel (RV) in a loop-type sodium-cooled fast reactor using a general-purpose finite element analysis code, FINAS/STAR, to understand its deformation behavior under extremely high temperature conditions and to identify the areas which should be focused to mitigate impacts of failure. The RV was heated from the normal operation condition to the sodium boiling temperature in the upper sodium plenum during 20 hours assuming depressurization. The analysis has revealed less significant stress and strain which were sufficiently lower than failure criteria. The upper body of RV was identified as the important area in terms of mitigation of structural failure. The RV was eventually deformed downward about 16 cm, resulting in no failure. This effect contributes to maintaining RV sodium level in a long term, thereby enhancing the RV resilience.

Development of safety design criteria and safety design guidelines for Generation IV sodium-cooled fast reactors

Futagami, Satoshi; Kubo, Shigenobu; Sofu, T.*; Ammirabile, L.*; Gauthe, P.*

Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs (TIC 2022) (Internet), 10 Pages, 2022/10

Three-dimensional structural analysis for enhancing resilience of next-generation nuclear structures under extremely high temperature conditions

Futagami, Satoshi; Ando, Masanori; Yamano, Hidemasa

Transactions of 26th International Conference on Structural Mechanics in Reactor Technology (SMiRT-26) (Internet), 9 Pages, 2022/07

Development of the pump-integrated intermediate heat exchanger in advanced loop-type sodium-cooled fast reactor for demonstration

Amano, Katsunori; Enuma, Yasuhiro; Futagami, Satoshi; Inoue, Tomoyuki*; Watanabe, Sota*

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 7 Pages, 2016/06

http://doi.org/10.1115/ICONE24-60064

In the framework of GIF, SDC and SDG for the generation IV SFRs have been developed in the circumstance of worldwide deployment of SFRs. JAEA and MFBR have been investigating design study of an advanced loop-type SFR to satisfy SDC in the feasibility study of SDG for SFR. In this study, the ability of the pump/IHX in the advanced loop-type SFR for the safety measures has been evaluated. In addition to the safety measures, maintainability and reparability are taken into account in the advanced loop-type SFR design study. The pump/IHX has been modified to satisfy these requirements. This paper describes the modifications for the ability to withstand a severe earthquake, the reliability of the guard vessel in the primary coolant leak, and the reliability of expansion joints in a sodium-water reaction. The evaluations of thermal transient, structural vibration with pump rotation and wear-out of IHX tubes, that has been adversely effected by the modifications, were described as well.

Conceptual design study for the demonstration reactor of JSFR, 5; Reactor cooling system design

Kisohara, Naoyuki; Ishikawa, Hiroyasu; Futagami, Satoshi; Xu, Y.*; Shimoji, Kuniyuki*; Kawamura, Masaya*

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 9 Pages, 2011/10

The cooling system of the JSFR adopts an integrated primary sodium pump/intermediate heat exchanger (IHX), dual structure straight tube steam generator (SG) and short elbow sodium piping layout. Since, however, this is the first experience applying these technologies to SFRs in Japan, design approaches have been evaluated and R&D has been undertaken. This paper addresses the design study of the cooling system of the demonstration reactor JSFR in terms of thermal-hydraulic and structural integrity. Recent studies have shown that these new technologies have potential to be applied to the JSFR.

Steam water pressure drop under 15 MPa

Liu, W.; Tamai, Hidesada; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki

Journal of Power and Energy Systems (Internet), 5(3), p.229 - 240, 2011/04

https://doi.org/10.1299/jpes.5.229

For a steam generator with straight double-walled heat transfer tubes that used in a sodium cooled faster breeder reactor, clarification of flow instability in heat transfer tubes is one of the most important research themes. As the first step of the research, thermal hydraulics experiments with water were performed under high pressure condition in JAEA with using a circular tube. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper summarizes the pressure drop characteristics under 15MPa. Several two-phase flow multiplier models were checked and then, it was found that both two-phase flow multiplier models of Chisholm and homogeneous can predict the present experimental data in high accuracy.

Thermal-hydraulic experiments under high pressure condition

Liu, W.; Tamai, Hidesada; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki

Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 9 Pages, 2010/05

https://doi.org/10.1115/ICONE18-30167

For steam generator with straight double-walled heat transfer tube that used in sodium cooled Faster Breeder Reactor, flow instability is one of the most important issues need researching. As the first step of the research, thermal hydraulics experiments with water were performed under high pressure condition in JAEA with using a circular tube with a similar inner diameter as that in the designed SG. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper summarizes the pressure drop characteristics under 15 MPa. Six models for the prediction of two-phase multiplier were evaluated. The results showed the Chisholm correlation and homogeneous model gave best predictions. Note that in the homogeneous model verification, the homogeneous model was only used in the friction loss calculation. In the calculation of void fraction, which is necessary for static head, drift flux model, instead of homogeneous model, was used.

A Study on LMFBR steam generator design without tube failure propagation in water leak events

Futagami, Satoshi; Hayafune, Hiroki; Fujimura, Ken; Sato, Mitsuru*

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9169_1 - 9169_8, 2009/05

The major target performance of the SG for commercialized FBR is not only economic performance but also property protection performance. The straight double wall tube SG is one of the SG candidates for commercialized FBR, and other SG concepts were studied in this paper. In proposing an alternative SG, alternative technological measures with a double wall tube were investigated and included reinforcing the tube against wastage and quick detection of initial tube leaks. Alternative SG concept candidates for preventing tube failure propagation and mitigation of water leak accidents were proposed through a combination of technological measures. At the end of JFY 2010, the straight double wall tube SG will be decided upon as the result of R&D activities, and alternative SGs evaluated in feasibility studies. A plan for studying feasibility with the technological issues of the alternative SG was proposed.

Applicability study of design options to prevent tube failure propagation on LMFBR steam generator

Futagami, Satoshi; Hayafune, Hiroki; Sato, Mitsuru*

JAEA-Research 2008-075, 94 Pages, 2008/10

JAEA-Research-2008-075.pdf:5.91MB

Based on the comprehensive evaluation in FS, straight double-wall tube SG was recommended as reference design. But straight double-wall tube SG has many technical issues. Alternative SG was designed by the objects as shown bellow. "safety performance is equal to reference design", "property protection performance is equal to reference design", "economically acceptable", "based on common technology with reference design", To design alternative SG with prevention of tube failure propagation, design options to resist the wastage by sodium-water reaction were evaluated from the point of view of property protection performance. And, candidates for alternative SG were proposed with combination of design options. By comparative evaluation of alternative SG by view of technical issues and adoption time, "Protective wall tube SG", "Coated wall tube SG" with resistance to tube wastage, and "Short length double wall tube SG" with low difficulty were elected.