FUKUSHIMA FIRST HOURS
by Dean Wilkie
In the aftermath of the massive 9.0 magnitude earthquake that occurred in Japan on 11 March, 2011, thousands of people were killed, thousands injured, and thousands were evacuated or relocated. The earthquake damaged the industrial infrastructure, affecting not only exports, but essential imports to the country. One of the most critical impacts was on the operating nuclear reactors which are used extensively in Japan to provide electrical power to the nation.
Those reactors worst hit were along the coastal areas where the magnitude 9.0 earth quake caused extensive damage. The earthquake rendered some critical equipment inoperable and weakened the facilities structures and system components. It has been estimated, but not proved, that piping connections to the reactor vessel and other systems were damaged resulting in leakage. Inspection of the interior of the facilities was unable to be performed due to the containment being isolated.
All of the reactors automatically shutdown as a result of seismic reactor safety system trips and several were impacted by initial damage done from the earth quake. The safety rods inserted within 4 seconds and the control rods were driven in within 5 minutes.
Following the facility shutdown from the earthquake, operators were in the process of reacting to the damage to critical systems and external electrical power loss while trying to place the reactors in a safe shutdown mode of operation. Reactor control room annunciation boards were showing many abnormal alarms and automatic system interlocks were trying to perform as designed in response to an accident of this magnitude.
As the facilities were responding, a massive tsunami, formed by the earthquake, was rapidly approaching the coastal areas. Once the tsunami reached shore it consumed everything in its path, destroying buildings not designed to withstand the destructive force. The tsunami, which was on the order of 5 times larger than ever anticipated struck the coastal reactor facilities and resulted in extensive damage.
The reactors of primary interest for this report were those at the Fukushima Daiichi No1 units 1, 2, 3, 4. Facilities. Reactors 1,2, and 3 were operating at the time of the earthquake and tsunami and No.4 was in a scheduled maintenance outage and was shutdown. The Fukushima reactors were operating on some of their diesel power or battery backed emergency systems due to loss of normal off and on site power. When the tsunami struck the plants, the resulting damage placed the facilities in a “Station Blackout” condition which is one of the beyond design basis accidents which can result in severe damage to the reactor cores and spent fuel storage pools. Without emergency core and spent fuel coolant and subsequent to the loss of power, explosions occurred at reactors 1, 2, 3, and 4 which caused extensive damage to the facilities. This likely resulted in breaches to the reactor vessels, concrete containment and building structure.
This report will try to focus on the first 2-3 hours starting from the earthquake to piece together what did/could have happened at the #1 Daiichi plant. Every attempt will be made to provide the references to help maintain factual accuracy and where no references exist the best knowledge base will be used and identified.
One point of interest, when the EQ happened at Fukushima , it put out of service an off site transformer station which cut out the connection to the electrical grid. This condition lasted for approximately 50 minutes before being restored and power inputs to the facilities had to be performed by electrical utility workers.
In addition, after the SCRAM the main steam turbine isolation valves shut which isolated steam to the main turbine and caused it to trip. At this point the electrical facility power that is generated by the turbine and used by various plant systems and instrumentation was also lost. One of the diesel generators was not available which left only 1 emergency diesel generator for the plant emergency systems. Reports were that this generator ran for approximately 19 minutes and was then lost. The total load of emergency components to protect the reactor was then left to survive on a short lived battery back up power system.
At first TEPCO reported that the operators were establishing emergency core cooling but later (nearly 2 months) reported that operators shut the emergency core cooling system down due to reasoned concerns that it may damage the reactor core. This left the reactor in a total station black out condition and provided no means of intervention to reduce the severity of accident. The only mechanisms at this point were actions by non human intervention like automatic relieving pressure with safety relief valves (and the electrical interconnections were not available to stop/close valves once opened) .
Data from the first moments and hour following the EQ has not been forthcoming from TEPCO. Some data released over 2 months later has shown recorder chart paper which had clearly been cut and put together (due to obvious different chart paper scales). In addition, released parameters for the plant changed/or did not give a complete enough picture as to what the extent of the transient condition was. Consideration must be given to the operators of plants who were very definitely trying to respond to a barrage of alarms, failures, annunciations flooding the control panels, computer failures etc.. The magnitude of this accident, similar to other major accidents like Chernobyl in Russia or Browns Ferry in the USA is not something that is trained for nor accurately modeled. The result led to the operators trying to regain control and prevent a multitude of failures from happening with no written procedure(s).
There have been many speculations, theories or scenarios identified on what really happened at Daiichi unit #1. However, without visual inspections, accurate reconstruction of the accident based on dependable and accurate data and independent review, all theories are just that.. theories. The assumptions in this report are based on a culmination of reports and theories on what happens to a BWR during a Loss of Coolant Accident combined with a total station black out and assumed 3 cases which identify the availability of post accident safety systems. The table at the end of this report explains those conditions.
FUKUSHIMA REACTOR #1
The unit 1 reactor is a boiling water reactor rated at 460MWe and was placed in service in 1971. The reactor pressure vessel is a MARK-1 designed by General Electric in the USA. The following is a summary of the unit 1 reactor information:
*PCV = Primary Containment Vessel
*RPV = Reactor Pressure Vessel
PCV Model Mark-1
Electric Output (MWe) 460
Max. Pressure of RPV 8.24MPa – 1195 PSI
Max. Temp of the RPV 300 C- 572 F
Max. Pressure of the CV 0.43MPa
Max. Temp of the CV 140℃- 284 F
Commercial Operation 1971
Emergency DG 2 – (water cooled)
Electric Grid 275kV
Plant Status on Mar. 11 In Operation
|Estimated total fuel in core||1000kg|
|After 3 years in core|
Fission products (~100 radioisotopes including Ce-137, I-131, Sr-90)
NORMAL OPERATION FLOW SYSTEMS
- Turbine inlet valves open
- Turbine bypass valves shut
- Turbine condenser cooled with auxiliary system
- Main Primary Coolant Pump – 138 tonne circulating water system
- Recirculating pump – circulates a flow path from the downstream side of the internal JET pumps then back to the JET pump inlet
- Suppression Pool Torus – Connected to sphere with vent lines, vacuum breakers for reverse flow – unit 1 1750 tonne water
- High Pressure Coolant Injection Pump – 5000gpm, 150-1000 psi (1134t/h – 1-6.8 MPa) used to provide coolant to the RPV
- Steam Driven Feed Water Pump – 600 gpm, 150-1000 psi (138t/h – 1-6.8 MPa) used to provide coolant to the RPV when steam is available.
NORMAL SHUTDOWN – RESIDUAL HEAT REMOVAL
- Electrically driven feed water pumps – circulate water through core with suction at the turbine condenser
- Turbine isolation valves – closed
- Turbine bypass control valve open to bypass steam from reactor vessel to the turbine condenser
- Turbine Condenser cooling water removes energy from the decay heat
- Residual Heat Removal Pump – recirculates water from the reactor vessel through the residual heat removal heat exchange for cooling down the temperature within the vessel
- Accident Management “NORMAL” conditions
- Control reactivity with control rods/poison
- Maintain water inventory in reactor pressure vessel
- Keep core covered with cooling water
- Maintain fuel cladding integrity – “don’t generate H2”
- Keep pressure in the reactor vessel below failure pressure
- Keep pressure in containment vessel below failure pressure
- Cool suppression pool below boiling point
- Vent gases through suppression pool stack
- Cooling systems designed for post –accident heat removal and control
- Standby liquid control system – boron injection
- Emergency Core Isolation Cooling
- High Pressure Coolant Injection HPCI
- Reactor Core Isolation Cooling RCIC
- (RCIC pump can be steam driven)
- Automatic Containment Depressurization
- Electric Low Pressure Coolant Injection LPCI
- (system at low pressure 10900 gpm
- @20psig – 2478 tonne/h @136kPa)
- Core Spray
- Emergency Diesel Generators – Typical installation is 2 – 6MWe per generator set and usually at least 2 per reactor unit
- Backup Battery Power – Connected to inverters to generate AC power, used only to power key instruments and controls, advertised capacity for 8 hours operation
The accident at Fukushima was not modeled and procedures not written to handle the multiple failures of systems. The result of the magnitude 9.0 earth quake, the loss of a transfer station power system and subsequent site power blackout, systems not performing their emergency functions and last but not least the overwhelming activities in the reactor control rooms. In this kind of emergency the operators are left to their training and knowledge of the plant to ensure they adhere to the accident management principles mentioned above. In addition and at the same time, they need to meet the emergency reporting requirements to TEPCO and local authority personnel to take immediate actions. Commercial power plants, particularly BWR operation with boiling going on in the core to produce the steam to drive the turbines and other essential equipment. This delicate balance must be maintained as they protect the core fuel temperatures.
Researching this accident has been very difficult due to changing reports from TEPCO/reactor safety/government officials. Sampling of water to detect potential fissions within the core after the accident has been sporadic and reports on the scene at the facilities has been restricted and limited to the public. This report is based upon the best modeling of severe accidents that was performed at the Browns Ferry facility where a cable tray fire resulted in similar conditions. This was the closest accident that this writer chose to base his assumptions and comparison to the Fukushima unit 1 reactor. It should be pointed out that approximately 2 months after the accident the TEPCO management announced that the unit #1 operators shut off the ECCS with intentions of protecting the core. Many other actions so far not reported or manual overrides could have made the situation worse or temporarily prolonged the inevitable meltdowns.
Table 1 illustrates the sequences of events with a time line beginning at the point of establishing the station black out and then progressing forward. Reports indicated that one of the emergency diesel generators were not operational due to damage from the earthquake and subsequent power loss. The remaining diesel generator reportedly operated for approximately 10 minutes as mentioned earlier and then was not available. At the time of the earthquake, there was no information verifying availability of sea water for injection to cool the heat exchangers on the turbine or other auxiliary systems, however loss of the electrical power grid and supply power from the earthquake would have rendered them inoperable. The complexity and unprecedented nature of this accident categorizes it as a beyond design basis accident, in other words a total loss of off and on site electrical power with multiple equipment failures that were thought to be impossible.
Note: Initial Condition Assumed at Reactor SCRAM + 10 minutes to Station Blackout
All times are in minutes.
SCRAM = Emergency shut down of a nuclear reactor.
HPCI = High Pressure Coolant Injection
RCIC = Reactor Core Isolation Cooling
SORV = Safety Operated Relief Valve
|5/17/2011 ACCIDENT PROGRESSION IN MINUTES|
|Station Blackout||Station Blackout||Station Blackout|
|No HPCI/RCIC||Manual HPCI/RCIC||Normal HPCI/RCIC|
|NO SORV||Manual SORV||Normal SORV|
|Start core melt||57||396||388|
|Corium slumps to bottom head||78||453||419|
|RPV bottom head fails||142||543||515|
|Corium starts boil off of water @ concrete containment floor||142.5||543.5||515.5|
|Corium starts melting containment floor||162||544||580|
|Containment failure and leakage starts||185||596||601|
|Drywell and Wetwell electric penetrations start to leak at T=204C – 400F|
|Drywell and Wetwell electric penetrations decompose and blow out of containment at T=260C – 500|
Table 1 Accident Progression Timeline
1. The Crisis at Fukushima Dai-chi Nuclear Power Plant, 4-2011
2. The 2011 off the Pacific coast of Tohuku Pacific Earthquake and the seismic damage to the NPPS, 4-4-4—4-4-2011
3. Fukushima Earthquake and Tsunami Blackout Accident, 5-16-11