Fumiya Tanabe

Analyses of core melt and re-melt in the Fukushima Daiichi nuclear reactors

Analyses are performed of the first core melt behavior of the Unit 1, Unit 2 and Unit 3 reactors of Fukushima Daiichi Nuclear Power Station on 11–15 March 2011 as well as the re-melt (melt again) behavior in another chaotic period of 19–31 March 2011. Analyses are based on a measured data investigation and a simple model calculation.

Analysis of Core Melt Accident in Fukushima Daiichi-Unit 1 Nuclear Reactor

In order to obtain a profound understanding of the serious situation in Unit 1 and Unit 2/3 reactors of Fukushima Daiichi Nuclear Power Station (hereafter abbreviated as 1F1 and 1F2/3, respectively), which was directly caused by tsunami due to a huge earthquake on 11 March 2011, analyses of severe core damage are performed. In the present report, the analysis method and 1F1 analysis are described. The analysis is essentially based on the total energy balance in the core. In the analysis, the total energy vs. temperature curve is developed for each reactor, which is based on the estimated core materials inventory and material property data. Temperature and melt fraction are estimated by comparing the total energy curve with the total stored energy in the core material. The heat source is the decay heat of fission products and actinides together with reaction heat from the zirconium steam reaction.

The effect of the virtual mass force term on the numerical stability and efficiency of system calculations

Abstract A simplified virtual mass force term has been implemented into TRAC-PFI. The coefficient of the virtual mass force term was determined so as to obtain the reasonable sound speed in two-phase flows. Implementation was easily accomplished without changing the basic solution method of SETS. Sample calculations and a small-break LOCA calculation were performed. The virtual mass force term was found to stabilize basic equations and numerical calculations. It was also found that a large amount of CPU time could be saved in the system calculation.

Creation of Interface System for Nuclear Reactor Operation Practical Implication of Implementing EID Concept on Large Complex System

Present status of on going JAERIs research project aiming at empirical evaluation of the EID concept is described. The research project is proceeded with three consecutive steps: design and implementation of the interface system, verification, and validation of the interface on a full-scope reactor simulator. Following detailed analysis of system structure in a way of “walking-through” the operation procedure on the simulator, a new interface system was created on this. In the newly created interface system, sets of graphical display formats representing higher levels of information items and new navigation mechanism were introduced to cope with complex nature of reactor system. At present, verification work for this new interface system is being conducted.

A scenario of large amount of radioactive materials discharge to the air from the Unit 2 reactor in the Fukushima Daiichi NPP accident

Based on an analysis of the measured data with review of calculated results on the core melt accident, a scenario is investigated for large amount of radioactive materials discharge to the air from the Unit 2 reactor. The containment pressure suppression chamber (S/C) should have failed until the noon on 12 March 2011 only by seismic load due to the huge earthquake on 11 March or by combination of seismic deterioration and dynamic load due to steam flowing-in through safety relief valve. Opening of the two safety relief valves (SRVs) at 14 March 21:18 should have resulted in discharge of large amount of radioactive materials through the S/C breach with the measured air dose rate peak value of 3.130E-3Sv/h at 21:37 near the main gate of the site. The containment drywell (D/W) should have failed at 15 March 06:25, at the cable penetration seal due to high temperature caused by the fuel materials heating up on the floor of the D/W, which had flowed out from the reactor pressure vessel. Then large amount of radioactive materials should have been discharged through the D/W breach with the measured air dose rate peak value of 1.193E-2Sv/h at 15 March 9:00.

Development of AI-Based Simulation System for Man-Machine System Behavior in Accidental Situations of Nuclear Power Plant

A prototype version of a computer simulation system named JACOS (JAeri Cognitive Simulation system) has been developed at JAERI (Japan Atomic Energy Research Institute) to simulate the man-machine system behavior in which both the cognitive behavior of a human operator and the plant behavior affect each other. The objectives of this system development is to provide man-machine system analysts with detailed information on the cognitive process of an operator and the plant behavior affected by operators actions in accidental situations of an NPP (nuclear power plant). The simulation system consists of an operator model and a plant model which are coupled dynamically. The operator model simulates an operators cognitive behavior in accidental situations based on the decision ladder model of Rasmussen, and is implemented using the AI-techniques of the distributed cooperative inference method with the so-called blackboard architecture. Rule-based behavior is simulated using knowledge representation with If-Th...

Vapor generation model for flashing in the initial blowdown phase

Abstract A vapor generation model for flashing in the initial blowdown phase is proposed based on a wall nucleation theory and a bubble transport model. Comparisons are made between the proposed model and the TRAC-PF1 model by using the MINCS code through analyses of three blowdown experiments with different scales. The present model well predicts the pressure undershoot in the vessel, while the TRAC model can not predict this typical thermodynamic nonequilibrium phenomenon.

Thermal-Hydraulics in Uncovered Core of Light Water Reactor in Severe Core Damage Accident, (I): Development of Models

A computer code SEFDAN is developed for one-dimensional thermal-hydraulics in a partially uncovered core of a light water reactor during a severe core damage accident. The developed models include: 1. Froth level (or dry-out level) calculation 2. Transition and mixing between convection flow regimes in convective heat transfer 3. Radiant heat transfer between solid walls and flowing gas 4. Heat generation by zirconium-water reaction 5. Crucibilization effect of zirconium-oxide layer 6. Steam starvation effect on zirconium-water reaction. This code does not calculate motion of fuel rod material but predicts the beginning of relocation. The major affecting models, froth level calculation model, heat transfer model and crucibilization model, are verified through analyses of experiments. This code can be used for thermal hydraulic analysis of a severe accident and fuel damage experiment until significant material relocation occurs.

Thermal-Hydraulics in Uncovered Core of Light Water Reactor in Severe Core Damage Accident, (III): Analysis of Power Burst Facility Severe Fuel Damage 1-1 Test with SEFDAN Code

The SEFDAN is a computer program to analyze the one-dimensional thermal-hydraulics of a partially uncovered core of a light water reactor in a severe degraded-cooling event. In order to verify the code and to obtain better understanding of the severe core damage process, SEFDAN has been applied to analyses of the thermal response of fuel rods in the Power Burst Facility Severe Fuel Damage 1-1 Test. The calculated results are in good agreement with the experimental results. The analysis indicates that fuel cladding temperature in a portion of the lowest one third of the test bundle would have reached the melting point of the ZrO2 during a rapid temperature excursion driven by the zirconium-water reaction. The result is consistent with the result of metallographic examination. The crucibilization effect of the ZrO2 layer played an important role in the reaction. Steam starvation condition would have occurred in contrast to the situation of the Scoping Test of the same test series. Zirconium-water reaction o...