Taisuke Yonomoto

Insights from review and analysis of the Fukushima Dai-ichi accident

An unprecedented earthquake and tsunami struck the Fukushima Dai-ichi Nuclear Power Plants on 11 March 2011. Although extensive efforts have been continuing on investigations into the causes and consequences of the accident, and the Japanese Government has presented a comprehensive report on the accident in the IAEA Ministerial Conference held in June 2011, there is still much to be clarified on what happened during the accident and why. This article aims at identifying what should be clarified further about the progression of the accident at Units 1–3 through the review and analysis of information released from Tokyo Electric Power Company and government authorities. It also discusses the safety issues raised by the accident based on the insights gained, in order to contribute to establishing a new framework that pursues continuous improvement toward the highest standards of safety that can reasonably be achieved.

Core Heat Transfer Coefficients Immediately Downstream of the Rewetting Front during Anticipated Operational Occurrences for BWRs

A heat transfer coefficient (HTC) model was developed for the prediction of post-boiling transition (post-BT) behavior that might occur during anticipated operational occurrences (AOOs) for boiling water reactors (BWRs). The model development was based on measurements of heat transfer coefficient, liquid droplet deposition rate, and droplet concentration in our experiments conducted at high pressure. The model focused on the heat transfer near the rewetting front where the cooling by droplet deposition significantly affects the propagation behavior of a liquid film. The correlation by Sugawara was validated for the prediction of the deposition by using the experimental data. The model was also expressed as a function of the distance from the rewetting front to use in analytical models for the rewetting propagation. Both expressions of the present model successfully predicted our experimental data simulating the BWR thermal-hydraulic conditions.

In-pile Experiment in JMTR on the Radiation Induced Surface Activation (RISA) Effect on Flow-boiling Heat Transfer

In-pile flow-boiling experiments were performed to investigate the possible enhancement of heat transfer by the radiation induced surface activation (RISA) effect. The test section was a 2-mm diameter 100-mm long bore in a SUS-316L stainless steel block heated electrically. The test section, housed in an irradiation capsule, was inserted into one of the irradiation holes in the Japan Materials Testing Reactor (JMTR) of the Japan Atomic Energy Agency (JAEA). Quasi-steady state experiments were conducted before irradiation (out-of-pile and in-pile before reactor operation), during irradiation and after irradiation (in-pile), for the same boundary conditions using the same test section block. This approach allowed direct evaluation of the RISA effect through comparison of experimental data. Boiling curves were obtained up to the onset of dryout in an annular dispersed flow, for mass fluxes ranging from 180 to 630kg/(m2s) under a fixed pressure of 420 kPa. The critical heat flux obtained during and after irradiation indicated an about 17% increase, on average, from that before irradiation. Meanwhile, the wall superheat at subcritical heat fluxes generally became greater than that before irradiation.

Review of five investigation committees’ reports on the Fukushima Dai-ichi nuclear power plant severe accident: focusing on accident progression and causes1

On March 11, 2011, the Tohoku District-off the Pacific Ocean Earthquake and the subsequent tsunami resulted in the severe core damage at TEPCOs Fukushima Dai-ichi Nuclear Power Plant Units 1–3, involving hydrogen explosions at Units 1, 3, and 4 and the large release of radioactive materials to the environment. Four independent committees were established by the Japanese government, the Diet of Japan, the Rebuild Japan Initiative Foundation, and TEPCO to investigate the accident and published their respective reports. Also, the Nuclear and Industrial Safety Agency carried out an analysis of accident causes to obtain the lessons learned from the accident and made its report public. This article reviews the reports and clarifies the differences in their positions, from the technological point of view, focusing on the accident progression and causes. Moreover, the undiscussed issues are identified to provide insights useful for the near-term regulatory activities including accident investigation by the Nuclear Regulation Authority.

RANS and LES Analyses on a Density Stratified Layer Behavior of Multicomponent Gas by Buoyant Jet in a Small Vessel

The analysis on a density stratified layer consisting of multiple gases in the reactor containment vessel is important for the safety assessment of sever accidents. Computational Fluid Dynamics (CFD) code has a potential to clarify detailed stratification phenomena in the containment vessel. In this paper, CFD analyses were carried out in order to investigate the erosion of the stratified layer by a vertical buoyant jet injected from the bottom of a small vessel. Although the Reynolds-Averaged Navier-Stokes (RANS) model is commonly used in industrial applications, it is known that the RANS analyses tend to overpredict effects of turbulent mixing and stratification erosion for these phenomena. This study carried out the RANS and Large-Eddy simulations (LES) in order to understand the detailed phenomena of the stratification erosion in a containment vessel, and clarify the problems of the RANS analysis from the comparison. As a result, although both the RANS and LES models calculated the erosion, the erosion rates calculated by the RANS models were faster than that by the LES model. The calculated erosion behavior was qualitatively different: the LES analyses showed the vertical helium turbulent transport was enhanced only in the radial region directly affected by the impinging jet, while the RANS analyses indicated the occurrences of such transportation at all the radial locations. Although more detailed validation is required using appropriate experimental data, this difference among the calculated cases suggests the importance of the improvement of the turbulence models in order to accurately predict turbulence damping in the stratification layer.Copyright

A simple mass and heat balance model for estimating plant conditions during the Fukushima Dai-ichi NPP accident

A simple evaluation method for the analysis of thermal-hydraulic transients in reactor pressure vessel (RPV) and primary containment vessel (PCV) is proposed to support understanding the accident behaviors of the Fukushima Dai-ichi nuclear power plant (NPP). Since most of the measurements of the plants were unavailable especially in the early stage of the accident, and the accessibility to the plants had been limited by radiation, analytical investigation for the plant was required to understand the plant conditions such as the magnitude of the damages. In order to provide easy-to-use technical tools to support the analytical investigation, we developed a simplified analysis code, named “HOTCB”, based on total mass and heat balances in a lamped parameter system. The HOTCB code has capabilities to treat two-phase fluid including water, steam, and non-condensable gas in a wide range of temperatures up to highly superheated conditions, and to consider heat structures, i.e. heat capacities and heat transfer to the fluid. The code was provided to Tokyo Electric Power Company (TEPCO) and was practically used for the analysis on the accident. This paper provides the details of the code and simulations of Unit 1 and Unit 2 reactors of Fukushima Dai-ichi nuclear power plant (NPP) as examples to show the usefulness of the code.

Heat conduction analyses on rewetting front propagation during transients beyond anticipated operational occurrences for BWRs

ABSTRACT Our previous study investigated the rewetting behavior of dryout fuel surface during transients beyond anticipated operational occurrences for BWRs, which indicated the rewetting velocity was significantly affected by the precursory cooling defined as cooling immediately before rewetting. This study further investigated the previous experiments by conducting additional experimental and numerical heat conduction analyses to characterize the precursory cooling. For the characterization, the precursory cooling was first defined quantitatively based on evaluated heat transfer rates; the rewetting velocity was investigated as a function of the cladding temperature immediately before the onset of the precursory cooling. The results indicated that the propagation velocity appeared to be limited by the maximum heat transfer rate near the rewetting front. This limitation was consistent with results of the heat conduction analysis using heat transfer models for the precursory cooling expressed as a function of distance from the rewetting front, the maximum wetting temperature, and the heat transfer coefficients in the wetted region. This paper also discusses uncertainties in the evaluation of transient heat flux from the measured surface temperature, and technical issues requiring further investigation.

ROSA/LSTF Experiment on a PWR Station Blackout Transient With AM Measures and RELAP5 Post-Test Analysis

An experiment on a PWR station blackout transient with the TMLB’ scenario and accident management (AM) measures was conducted using the ROSA/large scale test facility (LSTF) at Japan Atomic Energy Agency under an assumption of non-condensable gas inflow to the primary system from accumulator (ACC) tanks. The AM measures proposed in this study are steam generator (SG) secondary-side depressurization by fully opening the safety valves in both SGs with the start of core uncovery and coolant injection into the secondary-side of both SGs at low pressures. The LSTF test revealed the primary pressure started to decrease when the SG primary-to-secondary heat removal resumed soon after the coolant injection into the SG secondary-side. The primary depressurization worsened due to the gas accumulation in the SG U-tubes after the completion of ACC coolant injection. The RELAP5 code predicted well the overall trend of the major phenomena observed in the LSTF test, and indicated remaining problems in the predictions of SG U-tube collapsed liquid level and primary mass flow rate after the gas ingress. The SG coolant injection flow rate was found to affect significantly the peak cladding temperature and the ACC actuation time through the RELAP5 sensitivity analyses.Copyright