Home Environment Reflections on the Fukushima Daiichi Nuclear Accident
Accident Progression for Units 1–3
The Modular Accident Analysis Program (MAAP) is a computer code used by nuclear utilities and various research organizations to simulate the progression of severe accidents in a light water reactor (LWR) . The MAAP code cannot completely replicate the Fukushima Daiichi accident at the present time because of incomplete understanding about actual mechanisms and what the data indicate. Yet, the simulation is useful for checking the correctness of our understanding about severe accidents and constructing an integrated view of the accident; the discrepancy between simulation results and measurements gives valuable clues for further investigation. In this section, a summary of the accident progression of Fukushima Daiichi Units 1–3 is shown based on results recently obtained by validation studies for the MAAP code by comparing the simulation results with measured data. In this section as well, the accident progression is described by focusing on reactor water level and RPV/PCV pressure.
Fission-product (FP) atoms tend to have many neutrons compared to stable isotopes and are relatively unstable. Therefore, FPs decay to stable isotopes while releasing some energy. This energy liberated from FP is called decay heat. In a nuclear reactor, continuous removal of the decay heat is required even after termination of the nuclear fission reactions.
If decay heat cannot be removed, the water level in the reactor core decreases due to boiling. While it is better to maintain high pressure in RPV for suffi steam supply, it becomes impossible to insert water into the reactor externally at a high-pressure condition. Therefore, the pressure should be decreased sooner or later, depending on what type of the low-pressure injection system it is equipped with.
During the early stage of an accident under the situation of loss of ultimate heat sink (LUHS), because there are no measures to release the energy contained in the reactor core, PCV pressure is considered to indicate the degree of accumulation of decay heat. After the core uncovering has started, the massive pressure increase indicates hydrogen accumulation in the core, and a high degree of generation of metal water reaction, because PCV of Boiling Water Reactor (BWR) Mark-I was designed to suppress by condensing the steam released from RPV. PCV venting is the only way to release the energy to the environment in such a situation; however, this means a break in the PCV boundary, which is designed to prevent FP release. Again, there is a problem in the use of a low-pressure water injection system under high PCV pressure, so the pressure must be decreased. For this depressurization actuation, PCV venting is important, as in case of failure of the venting attempt, massive fission product might be emitted to environment.
As a result of the analysis for Unit 1 by comparing simulation results by MAAP to actual measurements, Fig. 2.1 shows the reactor water level changes, while Figs. 2.2 and 2.3 show changes of the reactor pressure and PCV pressure, respectively. In these figures, MAAP simulation results are labeled as “(analysis).” In this section, accident progression for Unit 1 is described in accordance with the following accident chronology (Table 2.2).
In Unit 1, all the cooling capability was lost due to the tsunami. Therefore, Unit 1 fell into a severe condition within 3 or 4 h after the Earthquake. It was not until the next morning (March 12) that TEPCO could inject water into RPV. And then, PCV venting was conducted at 14:30 on March 12. After that, the hydrogen explosion occurred.
220.127.116.11 From the Earthquake to Tsunami Arrival
At Unit 1, two isolation condenser (IC) systems were automatically activated due to the reactor pressure increase following the scram caused by the Earthquake. After that, the two IC systems were manually shut down and then IC subsystem-A was started up. The reactor pressure was controlled by manually repeating the
Fig. 2.1 Reactor water level change for Unit 1
Fig. 2.2 Reactor pressure changes for Unit 1
Fig. 2.3 PCV pressure changes for Unit 1
Table 2.2 Chronological accident description for Unit 1
aTime from MAAP calculation
start-up and shutdown of IC subsystem-A to maintain the pressure at a certain level. Maneuvering actions such as the starting up of the suppression chamber (S/C) in the cooling mode of the containment cooling system (CCS) were also being taken in parallel for a cold shutdown of the reactor. At 15:37 on March 11, 2011, however, all AC power supplies were lost due to the tsunami, followed by the loss of DC power supply.
Regarding the influence of the Earthquake, the issue of the possibility of a loss-of-coolant accident (LOCA) caused by the Earthquake was examined as described in Attachment 1–3 of Ref. .
18.104.22.168 From the Tsunami Arrival to Reactor Water Level Decrease
All cooling capabilities, including the steam-driven cooling system as well as motor-operated pump, were lost due to loss of control power, and all displays of monitoring instruments and various display lamps in the Main Control Room went out due to the loss of all AC and DC power. Approximately from 16:42 to 17:00 on March 11, 2011, part of the DC power supply was temporarily recovered, allowing the reactor water level to be measured for a while, which helped to confirm that it had decreased from the earlier level before the arrival of the tsunami. The level observed (by the wide range water level indicator) at 16:56 on March 11 was at the top of active fuel (TAF) +2,130 mm and had not decreased yet to TAF, although it was continuing to decrease (Fig. 2.1).
The analysis results shown in Fig. 2.1 suggest that the reactor water level reached TAF at about 18:10 on March 11, and the core damage started at about 18:50 (fuel cladding temperatures reached about 1,200 °C).
Even if the fuel starts to be uncovered, steam cooling prevents it from conspicuous temperature rises as long as sufficient steam is supplied from below. While decrease of the amount of steam generation due to decrease of water level progresses, once fuel claddings can no longer be cooled by steam cooling and their temperatures reach about 1,200 °C, large amounts of hydrogen are generated by water-zirconium reactions and the energy released from their oxidation reactions further raises fuel temperatures.
The situation continued that the IC operation could not be confi When part of DC power supply was temporarily recovered, it was observed that the isolation valve outside the containment of IC subsystem-A was operable (the status display lamp was “Closed”). The shift operators took action to open the valve at 18:18 on March 11. The operators confi that the status display lamp changed from “Closed” to “Open,” and they heard the steam generating sounds and saw steam above the reactor building, but the amount of steam was limited and it stopped a while later. Due to the operators' confi that steam generation had stopped and concern about the water inventory left in the IC shell side tank, at 18:25 the operators closed the isolation valve outside the containment on the return pipe. At 21:30 the operators took action again to open the isolation valve outside the PCV and confi the steam generating sounds and saw steam above the reactor building.
22.214.171.124 From the Reactor Water Level Decrease to PCV Pressure Increase
Reactor pressure of 7.0 MPa[abs] was measured at 20:07 on March 11 (Fig. 2.2), and drywell (D/W) pressure of 0.6 MPa[abs] at about 23:50; on March 12, D/W pressure of 0.84 MPa[abs] was measured at 02:30 and reactor pressure of 0.9 MPa[abs] at 02:45 (Fig. 2.3). In the meantime, although the exact timing is unknown, it was observed that at a certain time after 20:00 on March 11, the PCV pressure showed a sharp rise and the reactor pressure decreased despite no depressurization actions. BWR with MARK-I PCV is designed to suppress pressure increase by condensation at the suppression pool by steam from the reactor. Therefore, the sharp pressure rise is considered to be caused by gas leakage to the drywell.
A scenario was assumed in the analysis that steam had leaked from in-core instrumentation dry tubes or main steam pipe flanges due to temperature rise in the vessel caused by overheating of uncovered fuel and fuel melting.
When the fuel range water level indicators recovered functionality at 21:19 on
March 11 due to the temporary power supply, they showed that TAF was located at
+200 mm, but the reactor water level indicators seemed to have already been defective. In this period, there would be no conceivable reason for an increase in water level because no water was injected to RPV. This detail is described in Attachment 1–2 of Ref. .
The meltdown accident progressed as follows: When heated to high temperatures, fuel melted down from the core to the lower plenum, and then further down to the bottom of the PCV by breaking through the reactor vessel.
126.96.36.199 From Containment Vessel Pressure Increase to Containment Venting Operation
At about 23:50 on March 11, the D/W pressure measured 0.6 MPa[abs]. Thereafter, the indicator continued displaying high values. At around 04:00 on March 12, the dose rate near the main gate of the NPS site started to show an upward trend, which may have resulted from radioactive materials leaked from Unit 1.
It is highly possible that the molten fuel dropped to the bottom of the reactor vessel and further to the bottom of the PCV before 19:04 on March 12, when fi engines started continuous water injection into the reactor. It is possible that the relocation of molten fuel to the PCV raised the PCV pressure and temperature even more. This scenario is related to the amount of the water injected by fi engines .
When the molten fuel cannot be sufficiently cooled, the concrete of the PCV floor is heated up above its melting point and core-concrete reactions start, which dissolve the concrete. The core-concrete reactions generate non-condensable gases such as hydrogen, carbon monoxide, etc., resulting in a large impact on the containment pressure change and radioactive release behavior. But it is unknown to what extent core-concrete reactions actually occurred at that moment.
The D/W pressure was being maintained at about 0.7–0.8 MPa[abs], after reaching 0.84 MPa[abs] at about 02:30 on March 12, until PCV venting was successful. This fact of constant PCV pressure gives a strong suggestion that the PCV was leaking, because the PCV pressure should rise; when steam is produced due to water injection, PCV temperature rises, and gases are generated by core-concrete reactions, etc.
Fresh water was injected by fire engines from about 04:00 to 14:53 on March 12. But, since the fire protection system and make-up water system used for water injection are separated from the interior of the plant, part of the injected water had gone to other systems and equipment, not to the reactor. The analysis could yield consistent results with actual measurement data for containment pressures by assuming that the injection had not been enough to flood the core region and that only a fairly small amount of water, compared to the actual amount of discharged water by the fire engines, had been injected to the reactor.
188.8.131.52 From the Containment Venting Operation to Reactor Building Explosion
Three times at 10:17, 10:23, and 10:24 on March 12 the operation to open the small S/C vent valve was carried out from the main control room. There was no visible response in the D/W pressure, while the dose rate near the main gate increased temporarily at 10:40. A while later, when a temporary air compressor was connected to open the large S/C vent valve and it was started up at about 14:00, an up-current of steam above the stack was observed by a live camera and the D/W pressure decreased from 14:30 until about 14:50. No dose rate increase was observed near the main gate and monitoring post-8 (MP-8).
After the opening operation of the large S/C vent valve, the D/W pressure decreased from 14:30 through about 14:50. Later at 15:36, hydrogen in the reactor building exploded and the roof and outer walls of the uppermost fl were damaged. It can be considered that hydrogen gas generated mainly by water-zirconium reactions, which leaked together with steam and finally reached the reactor building, resulted in the hydrogen explosion. But its leak path, volume, explosion
aspects, and ignition source are still unknown.
184.108.40.206 From the Reactor Building Explosion to March 18
At 19:04 on March 12 after the reactor-building explosion, seawater injection was started by fire engines.
Water injection to Unit 1 and Unit 3 was halted once at 01:10 on March 14, when the water source used for these two units was depleted. Water injection to Unit 3 was resumed at 03:20 under critical conditions, when the water source was partly recovered by using an additional water supply, but water injection to Unit 1 was delayed. Water injection to Unit 1 and Unit 3 was again halted with the hydrogen explosion at Unit 3. Water injection to Unit 1 was eventually interrupted from 01:10 to 20:00.
Meanwhile, almost the whole core of Unit 1 dropped down to the lower plenum and most of that part dropped further to the containment pedestal, according to the analysis. There are many unknown matters concerning the location of debris, and the final status of accident progression.
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