MAAP code analysis for the in-vessel phase of Fukushima-Daiichi Nuclear Power Station Unit 1 and comparison of the results among Units 1 to 3

Sato, Ikken; Yoshikawa, Shinji; Yamashita, Takuya; Shimomura, Kenta; Cibula, M.*; Mizokami, Shinya*

Nuclear Engineering and Design, 422, p.113088_1 - 113088_24, 2024/06

https://doi.org/10.1016/j.nucengdes.2024.113088

MAAP code analysis focusing on the fuel debris condition in the lower head of the pressure vessel in Fukushima-Daiichi Nuclear Power Station Unit 2

Sato, Ikken; Yoshikawa, Shinji; Yamashita, Takuya; Cibula, M.*; Mizokami, Shinya*

Nuclear Engineering and Design, 404, p.112205_1 - 112205_21, 2023/04

https://doi.org/10.1016/j.nucengdes.2023.112205

Based on updated knowledge from plant-internal investigations, experiments and model simulations until now, the in-vessel phase of Fukushima-Daiichi Nuclear Power Station Unit 2 was analyzed using the MAAP code. In Unit 2, it is considered that the core material enthalpy was relatively low when it relocated to the lower plenum of the pressure vessel, then, cooled by the coolant and solidified there. Although the MAAP code tended to underestimate the degree of core-material oxidation during the relocation, this probable underestimation was compensated for by an existing study that was considered more reliable, so that more realistic debris conditions in the lower plenum could be obtained. Basic validity of the former prediction of the Unit 2 accident progression behavior was confirmed and detailed boundary condition for the later phase was provided. This boundary condition should be utilized for future studies addressing debris reheating process leading to lower head failure and debris relocation toward the pedestal.

Evaluation of core material energy change during the in-vessel phase of Fukushima Daiichi Unit 3 based on observed pressure data utilizing GOTHIC code analysis

Sato, Ikken; Arai, Yuta*; Yoshikawa, Shinji

Journal of Nuclear Science and Technology, 58(4), p.434 - 460, 2021/04

https://doi.org/10.1080/00223131.2020.1862718

Inverse analysis of steam and hydrogen generation history of Fukushima Daiichi Nuclear Power Station unit 3

Yoshikawa, Shinji

JAEA-Research 2019-004, 32 Pages, 2019/09

JAEA-Research-2019-004.pdf:2.77MB

Steam and hydrogen generation history and gas leakage area are inversely evaluated by a thermal hydraulic analysis code GOTHIC. The analyzed period in the accident progression is from the arrival of reactor liquid level at the top of active fuel (TAF) until start of depressurization of reactor pressure vessel(RPV) by activation of automatic depressurization system(ADS). Based on the measured behaviors of the RPV and PCV pressures from 6:30 of March 13th until the ADS activation, some leakage from RPV to PCV is supposed during this period. The leakage path and area are inversely derived on plural possible accident scenarios. The leakage area are estimated to be no greater than 1 cm. This result suggests that the gas flow at the time of the main slumping would have been through S/C, where vapor condensation was effective, thus certain contribution of non-condensable gases like hydrogen seems necessary to explain the observed D/W pressure increase.

Thermal-hydraulic analysis of MONJU upper plenum under 40% rated power operational condition

Honda, Kei; Ohira, Hiroaki; Sotsu, Masutake; Yoshikawa, Shinji

Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 12 Pages, 2010/10

In this study, we calculated the thermal hydraulics of the upper plenum of MONJU by the detailed analysis model using commercial FVM code, FrontFlow/Red. The present analysis model simulates all structures with high resolution meshes. The 1st order upwind and 2nd order central difference scheme were applied to the advection and diffusion terms, respectively. And RNG - model was applied to turbulence modeling. These calculation results indicated that the structures installed in the plenum except for UIS did not affect largely to the temperature and velocity, the flow characteristics in the present results had similar tendencies with porous media approached applied to the UCS region and that the difference between the temperature measured in the UCS region and that of SA outlets is relatively small.

Depositional records of plutonium and Cs released from Nagasaki atomic bomb in sediment of Nishiyama reservoir at Nagasaki

Kokubu, Yoko; Yasuda, Kenichiro; Magara, Masaaki; Miyamoto, Yutaka; Sakurai, Satoshi; Usuda, Shigekazu; Yamazaki, Hideo*; Yoshikawa, Shusaku*; Nagaoka, Shinji*; Mitamura, Muneki*; et al.

Journal of Environmental Radioactivity, 99(1), p.211 - 217, 2008/01

https://doi.org/10.1016/j.jenvrad.2007.11.010

In a sediment core of Nishiyama reservoir at Nagasaki, depth profiles of Pu/Pu ratio, Pu and Cs concentrations were determined. Sediments containing plutonium and Cs, which were fallout deposited immediately after a detonation of Nagasaki atomic bomb, were identified in the core. Observed below the sediments were macroscopic charcoals, providing evidence for initial deposit of the fallout. This is the first entire depositional records of plutonium and Cs released from the Nagasaki atomic bomb together with those from atmospheric nuclear tests.

Fast reactor development, From Monju to next FBRs; Dedicated to the Japanese society for multiphase flow

Yoshikawa, Shinji

Nihon Konsoryu Gakkai Nenkai Koenkai 2005 Koen Rombunshu, p.311 - 312, 2005/08

None

Development of "FBR plant engineering system"

Minami, Masaki*; Sakata, Hideaki; Yoshikawa, Shinji; Yamada, Fumiaki

JNC TN4410 2005-001, 123 Pages, 2005/03

JNC-TN4410-2005-001.pdf:6.81MB

A software system for straightforward and quick conceptual studies and technical evaluations of fast breeder reactor plants has been developed, mainly targeting the Japanese Demonstration Fast Breeder Reactor Monju. The studies and evaluations by this system used to be limited within steady and nominal conditions, excluding influences by changing specification values in accidental conditions. In this fiscal year, a new software component has been included in the system for simplified evaluation of transient characteristics, which is an essential for complete design of an FBR plant. This new evaluation function enabled to detect specifications to vary over acceptable range in transient conditions, and to notify necessity for re-adjustment of steady state design specification values. This system was also utilized to generate a sensitivity survey program in order to evaluate appropriateness of Monju design. The appropriateness of Monju design was evaluated in two ways. The first one is to follow specification selecting sequence as the as-built Monju has actually been designed. The second one is hypothetical deviations of major specifications of Monju and observation of the influences on other specifications. As a result, Monju design was confirmed to be adequate from the view point of need to meet design limitations at the design stage of Monju. This system is expected to be systematically re-arranged, because the system has now considerably complicated configuration after additions of various programs for many objects. This arrangement will facilitate future contribution of this system to technical studies in order for development of FBRs.

Power Generation with Fe2 VAI modules using Sodium Heat Source

Yoshikawa, Shinji; YOSHIKAWA JNC, Shinji; Suzuki, Ryosuke*; kondo, koki*; Nakai, Satoshi*; NAKAI, Satoshi*

The 23rd International Conference on Termoelectric, 0 Pages, 2004/12

A prototype system for thermoelectric power generation using the Heusler Fe2 VAI alloy has been studied as a potential heat recovery system from a liquid metal cooled fast breeder reactor. The modules consisting of 180 pairs using Fe2VAi cast alloy were mouted on 1.8 m long of the outer surface of a stainless tube (O.D. 34mm) which is internally heated by flowing liquid sodium (484-670K, 2-8 1/min). The element has a shape of a long cuboid, where one small rectangular end was thermally contacted with the tube surface and the other end was directly cooled by forced air flow to form a counter-type heat exchanger. The internal electric resistances of the modules were nearly same, however, the electromotive force varied widely because of the difference of thermal resistance between the module and tube surface. The output power generated from this prototype system was evaluated to be about 4.2 W/m assuming that all the circumference of the pipe surface is covered by the elements.

Improvement of the MSG code for the MONJU Evaporators; Additional function of reverse flow calculation on water/steam model and animation for post processing

Toda, Shinichi; Yoshikawa, Shinji; Watanabe, Osamu*; Kishida, Masako*; Oketani, Kazuhiro*

JNC TN4400 2003-005, 106 Pages, 2003/05

JNC-TN4400-2003-005.pdf:3.82MB

The improved version of the MSG code(Multi-dimensional Thermal-hydroulic Analysis Code for Steam Generators) has been released. It has been carried out to improve based on the original version in order to calculate reverse flow on water/steam side, and to animate the post-processing data. To calculate reverse flow locally, modification to set pressure at each divided node point of water/steam region in the helical-coil heat tansfer tubes has been carried out. And the matrix solver has been also improved to treat a problem within practical calculation time against increasing the pressure points. In this case pressure and enthalpy have to be calculated simultaneously, however, it was found out that using the bloci-Jacobean method make a diagonal-dominant matrix, and solve the matriz efficiently with a relaxation method. As the result of calculations of a steady-state condition and a transient of SG blow down with manual tip operation, the improvement on calculation function of the MSG code was confirmed. And an animation function of temperature contour in the sodium shell side as a post procssing has been added. Since the animation is very effective to understand themal-hydraulic behavior on the sodium shell side of the SG, especially in case of transient condition, the analysis and evalution of the calculation results will be enabled to be more quickly and effectively.