EXHIBIT 3
Special Report

INPO 11-005
November 2011

Special Report on
the Nuclear Accident
at the Fukushima
Daiichi Nuclear
Power Station
Revision 0

OPEN DISTRIBUTION

Open Distribution: Copyright © 2011 by the Institute of Nuclear Power Operations. Not for sale or commercial use.
All other rights reserved.
NOTICE: This information was prepared in connection with work sponsored by the Institute of Nuclear PowerOperations (INPO). Neither INPO, INPO members, INPO participants, nor any person acting on behalf ofthem (a) makes any warranty or representation, expressed or implied, with respect to the accuracy,completeness, or usefulness of the information contained in this document, or that the use of any information,
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INPO 11-005
Table of Contents
SectionPage

1.0 Introduction .........................................................................................................................1
1.1 Site Description ............................................................................................................2
2.0 Executive Summary ............................................................................................................3
3.0 Overview of Event ...............................................................................................................5
3.1 Status of Units Just Before the Earthquake..................................................................5
3.2 Earthquake ...................................................................................................................5
3.3 Tsunami........................................................................................................................7
3.4 Loss of Power ..............................................................................................................8
3.5 Core Cooling ................................................................................................................9
3.6 Containment Pressure Control ...................................................................................10
3.7 Spent Fuel Pools and Dry Cask Storage ....................................................................12
3.8 Alternative Injection Sources .....................................................................................12
3.9 Working Conditions ...................................................................................................13
4.0 Unit-Specific Event Narrative ..........................................................................................14
4.1 Unit 1 Narrative .........................................................................................................14
4.2 Unit 2 Narrative .........................................................................................................21
4.3 Unit 3 Narrative .........................................................................................................28
4.4 Unit 4 Narrative .........................................................................................................33
4.5 Spent Fuel Pools and Dry Cask Storage ....................................................................35
5.0 Radiological Overview ......................................................................................................38
5.1 On-Site Capabilities ...................................................................................................38
5.2 Site Boundary and Off-Site Capabilities....................................................................39
5.3 On-Site Dose Rates and Protective Measures ............................................................39
5.4 Site Boundary and Off-Site Dose Rates and Protective Measures ............................41
6.0 References ..........................................................................................................................44
7.0 Additional Information ....................................................................................................45
7.1 Earthquake Design Basis ...........................................................................................45
7.2 Tsunami Design Basis................................................................................................47
7.3 Station Specifications.................................................................................................48
7.4 Fukushima Daiichi Simplified System Drawings/Descriptions ................................54
7.5 Organizational Structure and Staffing .......................................................................62
7.6 Operator Training.......................................................................................................63
7.7 Regulatory Structure ..................................................................................................65
7.8 Acronyms and Abbreviations ....................................................................................70i

INPO 11-005
8.0 Event Progression and Timeline ......................................................................................71
8.1 Unit 1 .........................................................................................................................71
8.2 Unit 2 .........................................................................................................................81
8.3 Unit 3 .........................................................................................................................91
LIST OF FIGURES
Figure 3.3-1 General Elevations and Inundation Level
Figure 4.4-1 Unit 4 Standby Gas Treatment System Hydrogen Flow Path
Figure 4.4-2 Unit 4 Standby Gas Treatment System Filter Survey Results
Figure 5.3-1 Site Dose Rates
Figure 5.4-1 Initial Off-Site Survey Results
Figure 7.1-1 Preliminary Seismic Data
Figure 7.4-1 Isolation Condensers on Unit 1
Figure 7.4-2 High Pressure Coolant Injection (HPCI) – Typical Arrangement, Units 13
Figure 7.4-3 Reactor Core Isolation Cooling (RCIC) – Typical Arrangement, Units 2 and 3
Figure 7.4-4 Overview of Unit 1 Containment Vent System
Figure 7.4-5 Overview of Units 2 and 3 Containment Vent System
Figure 7.4-6 Simplified Electrical System
Figure 7.4-7 Fukushima Daiichi Electrical Distribution Damage
Figure 7.4-8: Generic cross-section of a BWR4 with a Mark I containment (similar to
Fukushima Daiichi)
Figure 7.6-1 Operator Progression and Training Requirements
Figure 7.7-1 Nuclear Emergency Response Organization Relationships
LIST OF TABLES
Table 3.2-1 Observed and Design Basis Seismic Data
Table 4.5-1 Spent Fuel Pool Data
Table 7.3-1: Fukushima Daiichi Unit Information (U.S. units)converted from source
document provided in Table 7.3-2
Table 7.3-2: Fukushima Daiichi Unit Information (metric units)
Table 7.3-3: Fukushima Daiichi System Information (U.S. units)converted from source
document provided in Table 7.3-4
Table 7.3-4: Fukushima Daiichi System Information (metric units)ii

INPO 11-005
1.0Introduction
This report provides a narrative overview and timeline for the earthquake, tsunami, andsubsequent nuclear accident at Tokyo Electric Power Company’s (TEPCO) Fukushima
Daiichi Nuclear Power Station on March 11, 2011. The purpose of this report is toprovide an accurate, consolidated source of information regarding the sequence of events
that occurred in the first days of the accident. The information contained in this report
may be used for determining future U.S. and international industry corrective actions.
Although INPO describes certain events and actions taken, those descriptions are notintended to reflect any analysis or assessment of the decisions made by any individual orentity. This report does not assess or analyze the effectiveness of plant workers or others
involved in response to the event.
This report reflects the best available information, most of which was obtained fromdirect and ongoing interaction with TEPCO. It focuses on the first days of the event forunits 1, 2, and 3 and includes some information on units 4, 5, and 6 as well as the spent
fuel pools. Because of the extensive damage at the site, some details of the event remain
unknown or have not been confirmed. All times are provided as Japan Standard Time
(JST).
This report includes the following sections:Executive Summary – high-level description of event milestones
Overview of Event – short description of key event factors
Event Progression – unit-specific narrative of event progression
Radiological Effects – radiological information, including radiological releases
Additional Information – station design information, drawings, and supporting data
This report was created from information provided by TEPCO, the Japanese government,
the International Atomic Energy Agency (IAEA), and several Japanese nuclear and safety
organizations. Some of the data included logs from the TEPCO Emergency Response
Center, unit-specific parameter values and chart recorder indications, and personalaccounts of the accident and plant conditions. In some cases, specific questions wereaddressed through INPO employees working in the INPO Emergency Response Center inAtlanta or in the TEPCO offices in Tokyo. Specific sources used to gather information
are provided in the reference section of this report.
TEPCO personnel have reviewed the content of this report for accuracy, based on theircurrent understanding of the event. Furthermore, TEPCO assisted in developing the unitspecific timelines, provided in Section 8.0, as well as the design basis information insections 7.1 and 7.2.
TEPCO openly shared information with INPO, responded to questions in a timely
manner, and provided resources when available to support the generation of this report.

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INPO 11-005
The utility is working to share the facts of this event with the industry and with theJapanese public. Without TEPCO’s assistance, this report would not have been possible.
This report is based on information available to INPO as of November 1, 2011. INPO
has verified as much information as possible, but it makes no warranties as to theaccuracy or reliability of the information. The information in this report has not beenverified through independent or on-site observations. The values provided, such asreactor parameters and seismic intensity, are preliminary and may be revised as more
information becomes available. The effects containment conditions may have had onthese indications have not been validated. Numerical values that include tenths of a unit
are not intended to imply accuracy or precision, but rather are a result of the conversion
from metric to U.S. units.
This report has been designated as “open distribution” and is available to the generalpublic. The information contained herein was provided to INPO without the expectationof confidentiality, and the report does not contain INPO proprietary information. INPO
members and participants may reproduce this document for business use. This report is
copyrighted, and written permission is needed for organizations other than INPO
members and participants to reproduce the information.
Copies of this report have been provided to TEPCO, the World Association of Nuclear
Operators (WANO), and the United States Nuclear Regulatory Commission.
1.1

Site Description

Fukushima Daiichi was the first of three nuclear generating stations operated by TEPCO.
The station is located on an 860-acre site in the Fukushima prefecture, approximately 160
miles (260 km) from Tokyo, on the northeast coast of Japan. It was one of the largest
generating stations in the world, consisting of six boiling water reactors capable ofgenerating 5,480 MWe total.
The units are designed such that
units 1 and 2, 3 and 4, and 5 and 6
share common facilities andstructures, such as a shared control
room and turbine building. The
station also has a shared spent fuelpool and dry cask storage facility.
The shared spent fuel pool islocated on the inland side (west) of
Unit 4. The dry cask storage
facility is located between units 1
and 5 along the coast.
Refer to Section 7.0 for additional
station and design-specificinformation on each unit.
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INPO 11-005
2.0

Executive Summary
On March 11, 2011, at 1446 (JST), a severe earthquake measuring 9.0 on the Richter
Scale occurred 112 miles (180 km) off the coast of the Fukushima Daiichi Nuclear Power
Station. The earthquake was the largest Japan has ever experienced. It caused all of theoperating units (units 1, 2, and 3) to automatically scram on seismic reactor protection
system trips. The earthquake damaged breakers and distribution towers, causing a loss of
all off-site electrical power sources to the site. The emergency diesel generators
automatically started and provided AC power to emergency systems. Three minutes after
the earthquake, the Japan Meteorological Association issued a major tsunami warning,
indicating the potential for a tsunami at least 3 meters high. Station workers were
notified of the warning and evacuated to higher ground.
Forty-one minutes after the earthquake, at 1527, the first of a series of seven tsunamis
arrived at the site. The maximum tsunami height impacting the site was estimated to be
46 to 49 feet (14 to 15 meters). This exceeded the design basis tsunami height of 18.7
feet (5.7 meters) and was above the site grade levels of 32.8 feet (10 meters) at units 14.
All AC power was lost to units 14 by 1541 when a tsunami overwhelmed the site andflooded some of the emergency diesel generators and switchgear rooms. The seawater
intake structure was severely damaged and was rendered nonfunctional. All DC power
was lost on units 1 and 2, while some DC power from batteries remained available on
Unit 3. Four of the five emergency diesel generators on units 5 and 6 were inoperable
after the tsunami. One air-cooled emergency diesel generator on Unit 6 continued tofunction and supplied electrical power to Unit 6, and later to Unit 5, to maintain coolingto the reactor and spent fuel pool.
With no core cooling to remove decay heat, core damage may have begun on Unit 1 on
the day of the event. Steam-driven injection pumps were used to provide cooling water
to the reactors on units 2 and 3, but these pumps eventually stopped working; and all
cooling water to the reactors was lost until fire engines were used to restore water
injection. As a result of inadequate core cooling, fuel damage also occurred in units 2
and 3. Challenges in venting containments contributed to containment pressuresexceeding design pressure, which may have caused containment damage and leakage.
Hydrogen generated from the damaged fuel in the reactors accumulated in the reactor
buildingseither during venting operations or from other leaksand ignited, producing
explosions in the Unit 1 and Unit 3 reactor buildings and significantly complicating the
response. The hydrogen generated in Unit 3 may have migrated into the Unit 4 reactor
building, resulting in a subsequent explosion and damage. The loss of primary andsecondary containment integrity resulted in ground-level releases of radioactive material.
Following the explosion in Unit 4 and the abnormal indications on Unit 2 on the fourth
day of the event, the site superintendent directed that all nonessential personnel
temporarily evacuate, leaving approximately 70 people on site to manage the event.
During releases, dose rates as high as 1,193 millirem per hour (mrem/hr) (11.93 mSv/hr)
were measured at the site boundary, approximately 0.6 miles (1 km) from units 14. The

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INPO 11-005
windows for the emergency response center had to be covered with lead shielding to
reduce dose rates in the center. Organized off-site radiation surveys began on March 16.
Radiation levels off site at that time ranged from 0.1 mrem/hr (1 µSv/hr) to 20 mrem/hr(200 µSv/hr). Thirty-seven miles (60 km) northwest of the station, the dose rate was 0.8
mrem/hr (8 µSv/hr). Water and soil samples indicated the presence of strontium, iodine,and cesium. Food and water restrictions were implemented in some areas as a result ofradioactivity. People within the 12.4 miles (20 km) surrounding the station were
evacuated, and those living up to 18.6 miles (30 km) away were directed to shelter insidetheir homes as the releases of radioactive gases and materials increased as the event
progressed and more fuel damage occurred. Potassium iodide tablets and powder were
distributed to local governments beginning March 21. Because the evacuations had
already been completed, however, the potassium iodide was not issued to the population.
Radiation surveys of the on-site areas surrounding units 13 showed dose rates as high as
13 rem/hr (0.13 Sv/hr) in areas around units 2 and 3. More detailed surveys performed
over the following weeks discovered localized dose rates greater than 1,000 rem/hr (10Sv/hr) around equipment and debris outside units 1 and 3.
Some personnel who responded to the event received high doses of radiation. Twocontrol room operators received the highest dosesa calculated internal and external
dose of 67.8 rem (0.678 Sv) and 64.3 rem (0.643 Sv). The majority of dose received bythese workers was internal (85-87 percent). Potassium iodide was provided to somestation personnel on March 13. As of the end of March, approximately 100 workers hadreceived doses of greater than 10 rem (0.1 Sv).
The Fukushima event was rated as a level 7 event on the International Nuclear andRadiological Event (INES) scale. The Nuclear Safety Commission of Japan estimatedapproximately 17 million curies (6.3 E17 Bq) of iodine-131 equivalent radioactive
material was released into the air and 0.127 million curies (4.7 E15 Bq) into the sea
between March 11 and April 5. The 1986 accident at Unit 4 of the Chernobyl nuclear
power plant was the only other nuclear accident to have a level 7 INES rating. According
to the IAEA, the Chernobyl accident resulted in approximately 378.4 million curies (14E18 Bq) of radioactive material being released into the environment.1
The combination of the earthquake and tsunami caused considerable damage to theJapanese coast. According to the government of Japan’s report to the IAEA, almost
500,000 residential buildings were damaged or destroyed. There was considerable
damage to roads, railways, and public and industrial utilities. Approximately 4 million
homes lost electricity. The total area inundated by the tsunami was approximately 217
square miles (561 square km). As of October 7, 2011, the Japanese Red Cross reportsthat almost 16,000 people are confirmed dead, and almost 4,000 remain missing.
Approximately 90 percent of the deaths were reported to be caused by drowning.

1

Chernobyl’s Legacy: Health, Environmental and Socio-Economic Impacts. The Chernobyl Forum 2003-2005
Second Revision

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3.0

Overview of Event
3.1

Status of Units Just Before the Earthquake

Unit 1 was in operation at rated power output before the event, with all safety systems
and both emergency diesel generators operable. High pressure coolant injection (HPCI)and both isolation condensers were available and in standby. Reactor water level andpressure were normal for power operations. Unit 1 had been in operation since
September 27, 2010.
Unit 2 was in operation at rated power output before the event, with all safety systems
and both emergency diesel generators operable. High pressure coolant injection andreactor core isolation cooling (RCIC) were available and in standby. Reactor water leveland pressure were normal for power operations. Unit 2 had been in operation since
September 23, 2010.
Unit 3 was in operation at rated power output before the event, with all safety systems
and both emergency diesel generators operable. Both high pressure coolant injection and
RCIC were available and in standby. Reactor water level and pressure were normal for
power operations. The startup transformer was out of service for planned modification
work. Unit 3 had been in operation since November 18, 2010.
Unit 4 was shut down and had been in an outage since November 30, 2010. All the fuelwas in the spent fuel pool to facilitate reactor pressure vessel shroud work. The cavity
gate was installed, isolating the spent fuel pool from the upper pools. The 4A emergency
diesel generator was out of service for planned maintenance, with the 4B emergency
diesel generator operable and in standby.
Unit 5 had been shut down and in an outage since January 3, 2011. Fuel had been loaded
into the reactor and the reactor pressure vessel (RPV) reassembled. Reactor water level
was high, reactor coolant system temperature was 192.2°F (89°C), and reactor pressure
was 1,037 psig (7.15MPa gauge) to support RPV leak testing. Decay heat removal was
secured at 0744 in preparation for the leak testing. Both emergency diesel generators
were operable.
Unit 6 had been shut down and in an outage since August 14, 2010. Fuel had been
loaded into the reactor and the RPV reassembled. Reactor water level was normal, andreactor coolant system temperature was 78.8°F (26°C) with the reactor coolant systemdepressurized. Residual heat removal (RHR) system B was being used as needed for
decay heat removal. All three emergency diesel generators were operable.
3.2

Earthquake

On March 11, 2011 at 1446, a 9.0 magnitude earthquake occurred off the eastern coast ofJapan. The epicenter of the earthquake was 112 miles (180 km) from the Fukushima
Daiichi site and the hypocenter was 15 miles (24 km) under the Pacific Ocean. The

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earthquake lasted approximately three minutes and resulted in the Japanese coastline
subsiding an average of 2.6 feet (0.8 meters).
Observed Data (interim)

Maximum Response Acceleration(gal)

Maximum Response Acceleration
Against Basic Earthquake Ground
Motion (gal)Horizontal(N-S)

Horizontal(E-W)

VerticalHorizontal(N-S)

Horizontal(E-W)

Vertical

Unit 1

460

447

258

487

489

412

Unit 2

348

550

302

441

438

420

Unit 3

322

507

231

449

441

429

Unit 4

281

319

200

447

445

422

Unit 5

311

548

256

452

452

427

Unit 6

298

444

244

445

448

415

Observation Point(the lowest
basement of reactor
buildings)

Fukushima
Daiichi2

Scram Setpoint(gal)

Horizontal(E-W)

Vertical

135100

Table 3.2-1 Observed and Design Basis Seismic Data
A gal is a unit of acceleration (cm/sec2) expressing the strength of an earthquake's tremors. Observed data isinterim and may be revised following further analysis. (See Section 7.1 for seismic data.)

The peak acceleration measured at Fukushima Daiichi was 0.561g (550 gal) in thehorizontal direction and 0.308g (302 gal) in the vertical direction at Unit 2. This
exceeded the design basis acceleration of 0.447g (438 gal) in the horizontal direction.
The design basis maximum acceleration was also exceeded in units 3 and 5. According
to the government of Japan, the probability for exceeding the design basis acceleration
was in the range of 10-4 to 10-6 per reactor-year. The design basis maximum accelerationin the vertical direction was not exceeded in any of the units.
The ground motion exceeded the reactor protection system setpoints, causing automatic
scrams. The power lines connecting the site to the transmission grid were damaged
during the earthquake, resulting in a loss of all off-site power. The emergency diesel
generators started and loaded as expected in response to the loss of off-site power tosupply electrical power, with the exception of one emergency diesel generator on Unit 4,
which was out of service for planned maintenance. Feedwater and condensate pumps,
which are powered by nonvital AC sources, were not available because of the loss of AC
power.
As the shaking from the earthquake subsided, the operators began their scram response.
Reactor pressure, reactor water level, and containment pressure indications for units 1, 2,
and 3 appeared as expected following a scram and did not indicate any potential breachof the reactor coolant system (RCS) from the earthquake. However, no detailed
walkdowns or further investigation has been performed.
2“Seismic Ground Motion Due to Great East Japan Earthquake and Seismic Ground Motion Accounted for in
Seismic Safety Assessments,” provided by TEPCO

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INPO 11-005

TEPCO activated its Headquarters for Major Disaster Countermeasures (Corporate
Emergency Response Center) in Tokyo to assess damage from the earthquake and tosupport recovery efforts. The Station Emergency Response Center was activated on site
to respond to the event.
In the time between the earthquake and the first tsunami, multiple seismic eventssomewith magnitudes between 6.4 and 7.9occurred within 100 km of the initiating event.
3.3

TsunamiThe earthquake generated a series of seven tsunamis that arrived at the site starting at
1527, 41 minutes after the earthquake. The first wave was approximately 13 feet (4meters) high. The height of this wave did not exceed the site design basis tsunami of
18.7 feet (5.7 meters) and was mitigated by the breakwater. A second wave arrived at
1535; however, the wave height is unknown, because the tide gauge failed (maximum
indicated level of the gauge is 24.6 feet or 7.5 meters). At least one of the waves that
arrived at the station measured approximately 46 to 49 feet (14 to 15 meters) high basedon water level indications on the buildings.
The tsunami inundated the area surrounding units 14 to a depth of 13 to 16 feet above
grade (4 to 5 meters), causing extensive damage to site buildings and flooding of theturbine and reactor buildings. Intake structures at all six units were unavailable because
the tsunamis and debris heavily damaged the pumps, strainers, and equipment, and theflooding caused electrical faults. The damage resulted in a loss of the ultimate heat sink
for all units. The diesel generators operated for a short time; but by 1541, thecombination of a loss of cooling water, flooding of electrical switchgear, and flooding ofsome of the diesel generator rooms (located in the basement of the turbine buildings andnot designed to withstand flooding) caused a loss of all AC power on site for units 15.
(Refer to Figure 7.4-7.)
Figure 3.3-1 shows the general elevations (typical for units 14) and the approximate
inundation level. The grade level of units 14 is 32.8 feet (10 meters) and is 42.7 feet (13meters) at units 5 and 6 above mean sea level (commonly referred to as OP, for the levelin the Onahama Port). The intake structures were at an elevation of 13.1 feet (4 meters)
for all units.
Main Control Room
Battery Room
Switchgear
EDG Room46 ft (14m)
33 ft (10m)
13 ft (4m)Figure 3.3-1 General Elevations and Inundation Level
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3.4

Loss of PowerIn the control rooms, as plant equipment and distribution panels flooded, lighting
gradually faded and instruments began to fail. Station batteries, which were designed tolast for 8 hours, were lost when the flooding grounded or damaged DC distribution
systems. The loss of DC power resulted in a loss of all lighting in the units 1-2 control
room within 51 minutes after the scram. (Note: Units 1 and 2 share a common control
room, as do units 3 and 4.) Normal lighting in the units 3-4 control room was lost, andonly emergency lighting remained. Control room operators began checking to see whichindications were still available.
Three air-cooled emergency diesel generators (EDGs) had previously been installed at
the station as a modification (2B, 4B, and 6B EDGs). These EDGs had independent fuelsystems and were capable of providing power to vital AC systems following a complete
loss of the seawater ultimate heat sink. Furthermore, AC distribution system cross-ties
had been installed between units, which allowed power to be transferred among units 1-2
and 3-4 or between units 5-6 for both the 6.9-kV and 480-V distribution systems. The
air-cooled EDGs were located above grade, and some of them survived the tsunami. The
distribution systems for the Unit 2 and the Unit 4 air-cooled EDGs, which were locatedbelow grade, flooded and failed during the tsunami. The Unit 6 air-cooled EDG andportions of the electrical distribution system survived the tsunami and were used toreestablish cold shutdown on units 5 and 6. Figure 7.4-7 illustrates the damage to theelectrical distribution system caused by the tsunami.
When all AC power was lost, TEPCO personnel notified the government that an
emergency condition existed. TEPCO corporate offices and the Japanese government
arranged for delivery of portable electric generators to the Daiichi site. The generators
were located; however, damaged roads and congested traffic prevented the generatorsfrom reaching the site quickly. Helicopters were considered, but the generators were toolarge and heavy to carry. Ultimately, TEPCO was able to secure some mobile generatorsfrom the Tohoku Electric Power Company. These generators, along with some TEPCO
generators, began to arrive at the site late in the evening of March 11 and continued toarrive into the next morning.
The portable generators were limited in their effectiveness because they could not be
connected to the station electrical distribution system as a result of the extensive damage
the tsunami and flooding caused. Workers checked motors and switchgear in an attempt
to find usable equipment to support cooling the reactors. The testing revealed that theUnit 2 standby liquid control (SLC) pumps were not flooded or damaged.
Based on the inspection results, the first mobile generator was placed adjacent to Unit 2,
and workers began to lay temporary cables from the generator to the associateddistribution panel for the SLC pumps. The temporary power cables were approximately
4 inches (10 cm) in diameter and 656 feet (200 meters) long and weighed more than
1 ton. Forty employees began to run the cables through the debris and flooded areas.
The force of the tsunami had blown manhole covers off, resulting in unmarked openings
in the ground. Aftershocks and subsequent tsunami warnings further slowed progress. In
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INPO 11-005
spite of the challenges, the workers completed the task on Unit 2 and terminated thetemporary cable to the associated power panel on March 12 at 1530.
At 1536, an explosion occurred in the Unit 1 reactor building. This explosion was most
likely caused by the buildup of hydrogen that had been generated in the Unit 1 reactor
core and leaked into the reactor building. The explosion injured five workers, and debris
from the explosion struck and damaged the cables and mobile generator that had been
installed to provide power to the standby liquid control pumps. The debris also damaged
the hoses that had been staged to inject seawater into Unit 1 and Unit 2. Fieldwork wassuspended as workers were evacuated to the Emergency Response Center foraccountability. Some of the debris was also highly contaminated, resulting in elevateddose rates and contamination levels around the site. As a result, workers were now
required to wear additional protective clothing, and stay times in the field were limited.
The explosion significantly altered the response to the event and contributed tocomplications in stabilizing the units.
3.5

Core Cooling

Following the tsunami, Unit 1 lost all AC and DC power, control room lighting, control
board instrumentation, and all cooling and high-pressure makeup water to the reactor.
Operators had been cycling the A isolation condenser as needed to control reactor
pressure and had just removed the condenser from service when the flooding began.
Neither HPCI nor the isolation condensers were in service when DC power was lost.
Unit 1 had no injection or core cooling in service. While indicated reactor water level did
not lower below the top of active fuel until the morning of March 12, calculations basedon conservative estimates later revealed that the core may have uncovered as early as
three hours after the earthquake, and fuel damage might have commenced approximately
1.5 hours later.
TEPCO estimates that there was no injection into the Unit 1 reactor for 14 hours and 9
minutes after the isolation condenser was secured, approximately one hour after thereactor shut down. Conservative calculations indicate that most of the core may havebeen damaged, and some of the fuel may have relocated to the bottom head of the reactor
vessel, although this has not been confirmed. Core cooling was eventually establishedwhen reactor pressure lowered sufficiently and a fire engine was used to inject fresh
water, followed by seawater.
Units 2 and 3 were relatively stable, with RCIC in operation after the tsunami. However,injection was eventually lost on these units, resulting in core damage.
TEPCO estimates that there was no injection into the Unit 2 reactor for 6 hours and 29
minutes following the loss of RCIC, approximately 70 hours after shutdown. The corebegan to uncover at approximately 1630 on March 14, three days after the tsunami.
Inadequate core cooling resulted in subsequent fuel damage. Conservative calculations
indicate that some of the fuel may have relocated to the bottom head of the reactor vessel,
although this has not been confirmed. Core cooling was eventually established when afire engine was used to inject seawater.
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On Unit 3, following the loss of high pressure coolant injection at 0242 on March 13,
approximately 36 hours after shutdown, TEPCO estimates that there was no injection into
the reactor for 6 hours and 43 minutes. The core began to uncover at approximately 0400
the second day after the tsunami. Inadequate core cooling resulted in subsequent fueldamage. Conservative calculations indicate that some of the core may have relocated tothe bottom head of the reactor vessel, although this has not been confirmed. Core cooling
was eventually established when a fire engine was used to inject seawater.
Based on the given plant conditions, it is expected that boiling occurred in the reference
legs of the reactor water level instruments, resulting in nonconservative water levelindications. After the event, TEPCO confirmed that adverse conditions in the drywell
may have resulted in boiling of the reference legs, causing indicated water levels to behigher than actual levels for all three units.
In each of the three units, it is postulated that there is extensive damage with limited andlocalized melting of the fuel and internals and limited vessel damage. The lack of corecooling to compensate for decay heat resulted in excessive fuel temperatures andoxidation of the zirconium cladding. The oxidation of zirconium in a steam environment
will create significant additional heat from the exothermic reaction and large quantities ofhydrogen. This hydrogen contributed to the increases in containment pressure and to thesubsequent hydrogen explosions on units 1, 3, and 4.
3.6

Containment Pressure Control

During a severe accident, containment pressure must be controlled to prevent damage tothe containment and to help remove energy to facilitate injecting water into the reactor
with low-pressure systems. Without heat removal systems (no AC power and a loss ofultimate heat sink), containment pressure and temperature will increase as energy fromthe reactor is transferred to the containment via safety relief valves or systems such as
RCIC and HPCI.
The TEPCO severe accident procedures provide guidance for venting containment. The
guidance directs venting when containment pressure reaches the maximum operatingpressure if core damage has not occurred. If core damage has occurred, venting thecontainment will result in a radioactive release, so containment is not vented untilpressure approaches twice the maximum operating pressure. In this case, the Emergency
Response Center personnel could not verify the integrity of the core, and the associatedguidance was applied in the decision to vent Unit 1.
The severe accident procedures specify that the chief of the Emergency Response Center(site superintendent) shall determine if containment venting should be performed. The
site superintendent may solicit input and advice from station management when making
this decision. Although government permission is not specifically required before
containment is vented, government concurrence is desired.

10

INPO 11-005
In the case of Unit 1, the site superintendent informed the government of his intention tovent the containment. Following this, he received concurrence from government
agencies to vent containment following a press conference, which was planned for 0300
on March 12. The associated evacuations were reported as complete at 0903, and theoperators were directed to vent containment at 0904.
A review of the applicable procedures revealed that the accident management guidelines
do not specifically require evacuations to be completed before venting. The proceduresdo, however, require management to be knowledgeable about the status of evacuations
and to coordinate venting containment with local authorities. The procedure wording has
typically been interpreted as encouraging the operators to verify evacuations arecompleted before venting.
The first indication of increasing containment pressure was not available until 2350 on
the night of the event, when workers connected the temporary generatorwhich was
being used to provide some control room lightingto the containment pressure
instrument. The indication read 87 psi (600 kPa). By this point, access to the reactor
building had already been restricted because of high dose rates. The lack of available
containment pressure indications early in the event may have prevented the operatorsfrom recognizing the increasing pressure trend and taking action earlier in the event.
Unit 1 containment was not vented successfully until approximately 1430 on March 12.
Additional challenges occurred because of high dose rates and a lack of contingency
procedures for operating the vent system without power, as well as the lack of prestaged
equipment, such as an engine-driven air compressor.
The decision to complete evacuations before venting containment, and the subsequent
equipment and radiological challenges encountered as operators attempted to establish avent path, delayed injection of water into the Unit 1 reactor. At approximately 0230 on
March 12, as Unit 1 depressurized, pressure in the reactor and in containment equalized
at approximately 122 psia (0.84 MPa abs). This pressure is above the discharge pressureof the station fire pumps and fire engines. Once pressure had equalized, further
reductions in reactor pressure were not possible until containment pressure had lowered.
As a result, little to no injection was achieved until after the containment was vented
successfully, which occurred at approximately 1430 on March 12.
High containment pressures in Unit 1 contributed to the amount of time Unit 1 did nothave adequate core cooling. In units 1, 2, and 3, the extended duration of hightemperature and pressure conditions inside containment may have damaged the drywellhead seals, contributing to hydrogen leaks and the subsequent explosions. Containment
leakage also contributed to ground-level radiation releases from units 1, 2, and 3.
See figures 7.4-4 and 7.4-5 for simplified drawings of the containment vent systems.

11

INPO 11-005
3.7

Spent Fuel Pools and Dry Cask Storage

Fukushima Daiichi had spent fuel stored in pools at each unit, in a common spent fuelpool, and in on-site dry cask storage. Spent fuel pool cooling flow was lost for all spent
fuel pools following the loss of off-site power and was not immediately restored when theemergency diesel generators started. Unconfirmed reports were that sloshing of the waterin the spent fuel pools resulted in a loss of some water during the earthquake. The
explosion in the Unit 4 reactor building caused structural damage to the Unit 4 spent fuelpool, but it is not clear if the integrity of the pool liner was compromised.
Subsequent analysis and inspections performed by TEPCO personnel determined that the
spent fuel pool water levels did not drop below the top of fuel in any spent fuel pool andthat no significant fuel damage had occurred. Current investigation results indicate thatany potential fuel damage was likely caused by debris from the reactor building
explosions.
The dry cask storage building was damaged by the tsunami, and some of the casks were
wetted. An inspection confirmed that the casks were not damaged by the event.
3.8

Alternative Injection Sources

Fukushima Daiichi had three fire engines available that had been added to improve firefighting capabilities following the 2007 Niigata-Chuetsu-oki earthquake that had affected
the Kashiwazaki-Kariwa nuclear power station. These fire engines could also be used as
an alternative low-pressure water source for injecting into the reactors; however, one wasdamaged by the tsunami and a second could not reach units 14 because of earthquake
damage to the road. Only one fire engine was immediately available to support the
emergency response on units 14. This fire engine was blocked from accessing Unit 1 bya fuel oil tank that had been displaced into one road by the tsunami and by the inability ofworkers to open a deenergized security gate on the other road. Workers broke through agate between units 2 and 3, allowing the truck to access Unit 1.
Although modifications had been made previously to allow fire engines to inject water
into the core spray system, establishing injection was still difficult. The fire engine did
not have sufficient discharge pressure to overcome the elevation differences and reactor
pressure. To compensate for this, the truck loaded water at the fire protection tank, thendrove to the Unit 1 reactor building to inject into the fire protection system. This
operation was slowed by debris and because the route took the fire engine under abuilding that had partially collapsed.
After some trial and error, workers established continuous injection by routing a suction
hose from a fire protection tank to the truck, then discharging to the fire protection
system piping and into the reactor via an installed modification to the low pressure
coolant injection system. The fire protection tank, however, only had one hoseconnection. As a result, injection into the reactor had to be stopped each time the tankneeded to be refilled so another fire engine, now available, could attach a hose and fill the12

INPO 11-005
tank. Seawater injection was eventually switched to a flooded pit, then to the harbor
itself.
3.9

Working Conditions

In the days after the earthquake and tsunami, a group of TEPCO employees, members of
the Japan Self-Defense Force, and other volunteers worked to stabilize the damaged
reactors. This group worked through extremely adverse conditions to complete theirassigned tasks.
The workers faced multiple challenges and hazards. The tsunami had caused
considerable damage. Large areas of the site were flooded or littered with debris. The
force of the tsunami coming ashore had blown manhole covers off, leaving unguarded
openings. Without power available, much of the work was performed in complete
darkness. Passageways or rooms in some areas of the plant were dark and flooded.
Elevated dose rates challenged worker ability to perform tasks in the plant and in thefield. For some of the higher dose jobs, such as the attempt to open the suppression
chamber vent valve on Unit 1, operators volunteered to perform the taskregardless ofthe potential hazards.
Operators worked to restore or maintain cooling to the reactors, sometimes using
unconventional or unique methods. Some of the tasks that were accomplished were notbased on existing procedure guidance or formal training. The workers were placed in
conditions that were beyond the design basis for the station and had to rely on theirfundamental knowledge and creativity to recover indications or operate systems. Whilethese efforts were not always successful on the first attempt, workers continued theirefforts until the desired results were achieved.
Ongoing aftershocks and tsunami warnings further challenged the workers. As expectedfollowing a major earthquake, hundreds of aftershocks occurred in the days after the
initial event. Two of the aftershocks on March 11 were larger than magnitude 7.0.
After the explosion in Unit 1, radiological conditions continued to degrade, and workers
were subjected to elevated and continuously changing dose rates and contaminationlevels. Under the threat of subsequent explosions in other units, they continued theirefforts to stabilize the reactors.
Because of the tsunami and earthquake damage to the surrounding communities, little
outside assistance was initially available. Some workers lost their homes and families tothe earthquake and tsunami, yet continued to work. Many workers slept at thestationusually on the floor. Because of food shortages, workers were commonly only
provided with a biscuit for breakfast and a bowl of noodles for dinner. Some of theseworkers remain on site today, still working to keep the reactors cool and prevent the
spread of contamination.
Two operators were killed when they became trapped while performing inspections in the
Unit 4 turbine building when the tsunami inundated the site and flooded the building.
13

INPO 11-005
4.0

Unit-Specific Event Narrative
4.1

Unit 1 Narrative

On March 11 at 1446 (T=0), an earthquake caused a loss of off-site power and an
automatic reactor scram. All control rods inserted; and several actions occurred,
including a loss of feedwater and condensate and main steam isolation valve closures, as
expected because of the loss of off-site AC power. The emergency diesel generators
started and loaded in response to the loss of off-site power and supplied power to thesafety systems. While reactor water level initially dropped because of the collapsing
steam voids, reactor water level was within the normal band and the operators did notneed to initiate high pressure coolant injection (HPCI).
Six minutes after the scram (1452), the isolation condensers (ICs) automatically initiatedon increasing pressure in the reactor pressure vessel (RPV), resulting in a decrease inreactor pressure as cooler water circulated through the reactor core from the ICs.
At 1503 (T plus 17 minutes), operators recognized that Unit 1 was exceeding cooldown
rate limitations and manually removed the isolation condensers from service by closing
the cold leg return containment outboard isolation motor-operated valves (MO-3A and B)(see Figure 7.4-1). These actions were consistent with procedure limitations of notexceeding a 100oF/hr (55oC/hr) cooldown rate. Operators determined that only one IC
was needed to control reactor pressure between 870 and 1,015 psig (6 to 7 MPa gauge).
Operators cycled the A IC system by opening and closing the motor-operated valve (MO3A) to control reactor pressure. Chart recorders indicate that the operators manually
started and stopped the IC system three times between 1510 and 1534 as RPV pressure
cycled and that the A IC was removed from service at approximately 1534, just minutes
before the loss of all AC and DC power.
At 1527 (T plus 41 minutes), the first tsunami arrived at the station. The subsequenttsunamis flooded and damaged the intake structure; and by 1537 (T plus 51 minutes), thetsunami had begun to flood the turbine building basement. The flooding wetted orsubmerged the emergency diesel generators and the AC and DC distribution systems,
resulting in a gradual loss of all AC and DC power. Between 1537 and 1550, the loss of
power caused a loss of normal control room lighting, indications, and controls. The
indications for HPCI and the IC systems faded and went out. TEPCO made an
emergency declaration because of the loss of all AC power and notified the government
and associated authorities.
The isolation condenser was the only system available to cool the reactor; and without
DC power, this system needed to be operated locally. The IC also required a source of
makeup water to the condenser to continue to function beyond 8 hours. Without power,
this makeup water would have to be provided using a diesel-driven fire pump. However,
the operators did not immediately place the IC in service. As a result, Unit 1 had no
injection or core cooling in service.

14

INPO 11-005
At 1636, another emergency was declared because of the inability to determine reactor
water level and the status of injection into the core. Batteries and cables were taken to
the control room in an attempt to restore control board instruments, with work focused on
restoring reactor water level indication. With no core cooling and extensive damage tothe site, workers began to investigate methods for venting containment without power
and reviewing methods for injecting water into the reactor using the fire protection
system or fire engines. The diesel-driven fire pump was started and allowed to idle instandby, ready to provide injection into the reactor when needed. In complete darkness,
operators began to align the alternative water injection valves from the fire protection
system to the core spray system by manually opening the valves in the reactor building.
Injection could not be achieved, however, until after the RPV was depressurized below
100 psig (0.69 MPa gauge).
Instruments and indications periodically appeared to function but would soon lose power
and fail again. On one occasion, the valve position indications for the IC MO-2A and
MO-3A motor-operated valves began working. The operator noticed that the valves bothindicated closed. At 1818, the operator opened both valves using their main control room
switches in an attempt to place the isolation condenser in service. While some steam wasinitially seen coming from the condenser, the steam faded. At 1825, the operator closed
the MO-3A valve to remove the system from service. The reason for this action has notbeen determined. As a result, there was no cooling method aligned to remove decay heatfrom the reactor.
Because the control room had no working indications, operators checked reactor pressure
locally in the reactor building. At 2007, reactor pressure indicated 1,000 psig (6.9 MPa
gauge). Reactor water level was still unknown.
At 2049 (T plus 6.1 hours), workers restored some temporary control room lighting in theunits 1-2 control room when a small portable generator was installed.
At 2050 (T plus 6.1 hours), the Fukushima prefecture began to direct residents living
within 1.2 miles (2km) of the station to evacuate.
Water level indication was restored in the control room at 2119 (T plus 6.5 hours).
Indicated reactor water level was approximately 8 inches (200 mm) above the top of
active fuel (TAF).
At 2123 (T plus 6.6 hours), the Prime Minister expanded the evacuation zone to 1.9 miles(3 km), with residents living within a 6.2 mile radius (10 km) directed to take shelter.
Operators placed the A IC back in service at about 2130 (T plus 6.7 hours), when once
again the indications began to work. By this point, no cooling or injection had been
provided to the reactor for almost 6 hours, and core damage was most likely occurring.
While steam was observed coming from the condenser vent, it is not clear that the IC
went into service as expected. Inspections performed in September 2011 revealed that15

INPO 11-005
the A IC valves did open but the water level in the secondary side remained at 65 percent,indicating that the system may not have functioned as designed.
Dose rates in the reactor building increased to such a level that, by 2151 (T plus 7.1
hours), access to the building was restricted. By 2300 (T plus 8.2 hours), dose rates ashigh as 120 mrem/hr (1.2 mSv/hr) were detected outside the north reactor building
personnel air lock door. Dose rates in the control room also increased.
Just after midnight on March 12 (T plus 9.3 hours), the site superintendent directed theoperators to prepare to vent the primary containment vessel. In the control room,
operators assembled piping and instrumentation drawings, the accident management
procedures, valve drawings, and a white board. Because there was no procedure tooperate the vent valves without power, the operators began to develop a plan for venting,including how to operate the valves manually. They determined that both the motoroperated containment vent valve (MO-210) and the small suppression chamber airoperated vent valve (AO-90) could be operated manually (see Figure 7.4-4). At 2350 (Tplus 9.1 hours), containment pressure indicated 87 psia (0.6MPa abs), exceeding thecontainment design pressure of 62.1 psig (0.428 MPa gauge).
At 0130 (T plus 10.7 hours), TEPCO officials informed the Prime Minister, the Ministerof Economy, Trade and Industry, and the Nuclear and Industrial Safety Agency of plansto vent containment. All concurred with the venting of units 1 and 2 containments. The
government planned a 0300 press conference to announce the venting. The TEPCO
corporate Emergency Response Center instructed the station to vent the containments
following the press conference. The operators continued preparations to vent thecontainments.
At 0148 (T plus 11 hours), the installed diesel-driven fire pump that had been running instandby to inject into the reactor stopped operating. To restart the fire pump, workers
carried diesel fuel to the pump and refilled the fuel tank, but attempts to start the engine
depleted the batteries. Workers then retrieved spare batteries that were stored in an office
and attached them, but the engine still would not start.
In parallel, efforts to use a station fire engine as an injection source to the reactor wereongoing. The damage from the earthquake and tsunami made this task difficult. The
earthquake and tsunami also damaged fire hydrants and caused fire system leaks. While
workers were able to close valves and isolate the leaks, the damage made it impossible touse filtered water as a water source. Fire protection tanks remained available for use as awater source.
The station had three fire engines, but only one was available to support injecting water
into the Unit 1 reactor. One fire engine was damaged by the tsunami and was not
functional. The second was parked adjacent to units 5 and 6 but could not be driven to
Unit 1 because of earthquake damage to the road and debris from the tsunami. The
remaining fire engine, which was located near units 3 and 4, was functional. Workers
had to clear obstacles and debris to move the fire engine to Unit 1. A heavy fuel oil tank,16

INPO 11-005
which had been displaced by the tsunami, made one access road impassable. A security
gate that had lost power and would not open blocked another road that provided access toUnit 1. Workers broke a lock on the gate between units 2 and 3, allowing the fire engine
to arrive at Unit 1.
Alternatives for injecting water via fire protection lines were reviewed, and additionalfire engines and water transportation by the Japanese Self-Defense Force were also
considered.
By 0230 (T plus 11.7 hours), indicated containment pressure had increased to 122 psia
(0.84MPa abs), which is approximately twice design pressure. Indicated reactor pressure
decreased to 116 psig (0.8 MPa gauge), and reactor water level indicated 19.7 inches(500 mm) above TAF by the lowest indication. At this pressure, the containment had
equalized with reactor pressure and was still greater than the discharge pressure of thediesel-driven fire pump lined up to inject water into the reactor. There was no steam flowfrom the reactor to cool the fuel, and there was no source of injection into the reactor.
TEPCO is not sure how the Unit 1 reactor depressurized. Because reactor andcontainment pressure equalized, it appears the reactor depressurized because of either astuck-open relief valve or a breach of the reactor coolant system or reactor pressure
vessel. The isolation condenser may have helped with the pressure reduction; however,
had the isolation condenser caused the depressurization, reactor pressure likely would
have continued to decrease until the shell-side boiled dry. Then the reactor coolant
system would have repressurized and would not have equalized with containmentpressure.
At approximately 0300, a press conference was held to announce the venting of thecontainments. At the station, however, workers were not directed to perform theevolution, and indicated primary containment pressure remained well above the 62.1 psig
(0.428 MPa gauge) design pressure. Allowing the containment to be exposed topressures above design pressure may have caused containment penetrations and seals todegrade and leak, but this has not been verified. Indicated containment pressure began todecrease unexpectedly and stabilized near 113 psia (0.78 MPa abs) without venting.
As the morning progressed, plant conditions continued to degrade. In preparation for
venting the containment, workers attempted to enter the reactor building to performsurveys. When the reactor building air lock door was opened, the workers saw steam andclosed the door. No surveys were performed.
The first indications of an off-site release were detected at 0450 (T plus 14.1 hours) whena dose rate of 0.1 mrem/hr (1µSv/hr) was measured at the site boundary. The source of
this release has not been confirmed, but the timing correlates with an unexplained slow
reduction in containment pressure without venting. By 0500 (T plus 14.2 hours), workers
were directed to wear full face masks with charcoal filters and coveralls in both thecontrol room and in the field. Increasing dose rates in the Unit 1 control room caused
operators to periodically move to the Unit 2 side of the room where dose rates were17

INPO 11-005
lower. At 0514 (T plus 14.5 hours), workers noted an increase in radiation dose rates in
the plant concurrent with the decrease in containment pressure. Workers believed thismay have indicated a leak from the containment. This was reported to the government.
Over the next 30 minutes, radiation levels at the site boundary increased. At 0544 (T plus
15 hours), the Prime Minister expanded the evacuation zone to 6.2 miles (10 km).
As reactor pressure and containment pressure slowly decreased, a fire engine began
injecting fresh water from a fire protection water storage tank into the reactor via the corespray system. Although reactor pressure was not recorded, containment pressure wasapproximately 107 psia (0.74 MPa abs). The low discharge pressure of the fire pump
was only slightly higher than reactor pressure, so the injection flow rate was low.
Complications in maintaining the injection lineup further reduced injection rates.
Initially, the fire engine refilled with water at the fire protection water storage tank, thendrove close to the reactor building and injected water through a fire protection line
connected to a core spray line. This was because the fire protection water storage tank
was at a low elevation, and the workers were concerned about the discharge pressure of
the fire engine being insufficient to overcome reactor pressure and inject into the core.
Further complications, such as driving the fire engine under damaged buildings locatedbetween the tank and the unit, exacerbated these delays. During these first few hours, thecalculated fire engine injection rate was low, averaging less than 10 gpm (38 lpm).
After some trial and error, workers established continuous water injection from the fireengine. A hose was run from the suction of the fire engine to the fire protection water
storage tank, allowing the fire engine to discharge water directly into the fire protection
line system and into the reactor.
An additional fire engine arrived on site and was used to repeatedly transport fresh waterfrom the fire protection tank at Unit 3 to the fire protection tank at Unit 1. The Unit 1 fire
protection tank had only one hose connection; so to refill the tank, the fire engine thatwas injecting water into the reactor had to be disconnected from the tank. As a result,water injection into the reactor was stopped each time the second fire engine needed toreplenish the Unit 1 fire protection tank.
At 0650 (T plus 16.1 hours), the Minister of Economy, Trade and Industry ordered
TEPCO to vent the Unit 1 containment. TEPCO personnel, however, had just learned
that some residents inside the evacuation zone were not sure which direction to evacuate,
so they had not left yet. The Prime Minister arrived at the station at 0711. After somediscussion, TEPCO confirmed its plans to vent containment at 0900 after evacuations
were completed; and at 0804, the Prime Minister left the station. By this time, indicatedreactor water level had dropped below the top of active fuel by the lowest indicatedreading.
TEPCO informed the local governments that venting would start at approximately 0900.
Venting was being coordinated with the local governments in an attempt to ensure
evacuations were completed before venting commenced. Station procedures for venting
containment did not specifically state that evacuations be completed before venting. The

18

INPO 11-005
procedures stated that venting containment should be coordinated with local governments
and that the station should be knowledgeable about the status of evacuations. Thesestatements had been interpreted as providing guidance to verify evacuations werecompleted before venting.
The control room operators formed three teams to perform the venting, with two
operators on each team (one to perform actions and the other to assist by holding
flashlights and monitoring dose rates, as well as for other safety concerns, such as
ongoing aftershocks). Because there were no means of communicating with the fieldteams, they were dispatched one at a time, with the next team leaving only after thepreceding team returned.
In preparation for manual venting of the containment, a radiological evaluation ofworking conditions in the torus room was provided to the Emergency Response Center.
Based on radiation levels of 30 rem/hr (300 mSv/hr), workers were limited to 17 minutes
stay time to remain below the emergency response radiation exposure limit of 10 rem(100 mSv). Workers were required to wear a self-contained breathing apparatus with a
20-minute air supply and were given potassium iodide tablets.
At 0903 (T plus 18.2 hours), evacuations south of the plant were confirmed as being
completed, and the first team was dispatched to open the motor-operated containment
vent valve (MO-210) (see Figure 7.4-4). The team opened the valve the desired amount.
The operators received approximately 2.5 rem (25 mSv) each while performing the task.
The team returned to the control room; and by 0930, the second team had been
dispatched to open the air-operated suppression chamber vent valve in the torus room
(AO-90). To open this valve, the team would have to enter the torus room at one locationand travel to the other side of the room to manipulate the valve. The team was
unsuccessful, as dose rates in the torus room quickly exceeded their limits; and theoperators turned back. One of the operators received 10.6 rem (106 mSv), exceeding his
emergency dose limit of 10 rem (100 mSv).
The control room operators decided not to dispatch the third team because of the doses
received. They notified the Emergency Response Center (ERC) of the inability to open
the air-operated vent valve (AO-90). As a result, TEPCO personnel had to devise a new
method to open the air-operated valve. The ERC began working on methods to open thelarge suppression chamber air-operated vent valve (AO-72). This would require DC
power and a temporary air source. ERC personnel instructed the control room to attempt
to operate the small air-operated suppression chamber vent valve remotely, assuming
there would be sufficient residual air pressure in the system to operate the valve.
Workers continued their efforts to vent containment, while other groups worked to install
mobile generators and stage fire hoses to allow seawater injection into the reactor.
Workers continued to inject fresh water using a fire engine.
At 1017 (T plus 19.5 hours), workers had installed temporary batteries to provide DC
power to the small air-operated suppression chamber vent valve (AO-90). Operators

19

INPO 11-005
attempted to open the valve from the control room, relying on residual air pressure in the
instrumentation air system. Operators made three attempts to open the small air-operated
valve (at 1017, 1023, and 1024).
At 1040 (T plus 19.9 hours), radiation levels increased at the main gate and themonitoring post. Workers initially believed the radiation levels indicated the small airoperated suppression chamber vent valve (AO-90) had opened. However, by 1115 theradiation levels were decreasing and indicated containment pressure remained high,indicating that the venting was not fully effective. While this has not been confirmed, thetrend in radiation levels indicates that the small air-operated vent valve may have opened
intermittently, and this may have resulted in some of the downstream system pressurizing
and gases leaking from the system. However, it is postulated that the valve did not
remain open long enough to allow the pressure to blow the rupture disk and vent thecontainment through the ventilation stack.
The ERC was informed that a small air compressor was available in a subcontractor’s
office. Workers retrieved drawings and took pictures of the connection point and planned
how to install the compressor to enable remote operation of the large air-operated
suppression chamber vent valve (AO-72) from the control room. The temporary air
compressor was located and transferred to the reactor building equipment bay. By 1400
(T plus 23.2 hours), the compressor was installed and started. At 1430almost 24 hours
after the event had begunthe rupture disk opened and containment venting commenced.
Containment pressure began to decrease, and the injection flow rate of water into thereactor subsequently increased. Calculations, based on the total volume of water injected
into the reactor, demonstrate that the injection rate was approximately 50 gpm (189.3lpm). By 1453, the Unit 1 fire protection tank was running out of water, so the site
superintendent directed the injection of seawater into the reactor. Water transfers fromother sources continued, while workers staged hoses and prepared to inject seawater into
the reactor. Work to install a temporary generator, which would provide power to theUnit 2 standby liquid control and control rod drive pumps, was nearing completion. Thispower could be cross-tied to the Unit 1 systems, providing injection sources in Unit 1 aswell.
At the same time, hydrogen generated from the high-temperature interaction of zirconium
and steam was being released from the reactor into the containment. Some of these gases
found their way into the reactor building through leaks, most probably in the primarycontainment vessel penetrations because of the excessive pressure. Other potential leakpaths included possible damage to the hardened vent or backflow through the unit
standby gas treatment system into the reactor building; however, the exact leak path has
not been determined. As the gases accumulated in the reactor building, an explosive
concentration of hydrogen developed, resulting in an explosion at 1536 on March 12.
The explosion breached the reactor building and allowed radioactive materials to bereleased to the environment; it also damaged temporary power cables, generators, fireengines, and the fire hoses that had been staged to inject seawater. The five workers who
were injured by the blast were carried to safety. The remaining workers evacuated to the20INPO 11-005
ERC for accountability. Additionally, the debris spread by the explosion was highly
radioactive, further complicating the event response. The explosion also damaged thedoor to the control room, which had been opened to allow workers to install temporary
power cables. The open door allowed airborne radioactive material to enter the control
room. All injection into the core was lost.
Less than an hour after the explosion, radiation dose rates at a station monitoring post
along the site boundary had reached 101.5 mrem/hr (1,015 µSv/hr). By 1825, the Prime
Minister had expanded the evacuation zone to 12.4 miles (20 km).
The operators lined up a fire engine to inject seawater into the reactor through the corespray system and commenced injecting seawater at 1904 on March 12. Boron was thenadded to the water source to address criticality concerns.
This situation continued over the next several days as site personnel attempted to restore
electrical power to the unit. Off-site power was restored to Unit 1 on March 20, ninedays after the event.
4.2

Unit 2 Narrative

On March 11, 2011 at 1446 (T=0), an earthquake caused a loss of off-site power and an
automatic reactor scram. All control rods inserted and several actions occurred, including
a loss of feedwater and condensate and main steam isolation valve closures, as expected
because of the loss of off-site AC power. The emergency diesel generators started andloaded in response to the loss of off-site power and supplied power to the safety systems.
Reactor water level initially dropped because of the collapsing steam voids, as expected;
and operators initiated reactor core isolation cooling (RCIC) to maintain reactor waterlevel after the scram. One minute later, RCIC automatically shut down because of a highreactor water level. The operators waited approximately 10 minutes for level to lower
and then restarted RCIC. Torus cooling and spray were started to cool the suppression
chamber pool, removing the heat introduced by the RCIC turbine exhaust.
At 1527 (T plus 41 minutes), the first of a series of seven tsunamis generated by theearthquake arrived at the station. One minute later, RCIC again shut down because of ahigh reactor water level. Operators waited for reactor water level to lower, then restarted
RCIC.
The subsequent tsunamis flooded and damaged the intake structure; and by 1541 (T plus
55 minutes), the tsunami had begun to cause flooding in the turbine building basement.
The flooding wetted or submerged the Unit 2A emergency diesel generator and the AC
and DC distribution systems, resulting in a gradual loss of all AC and most DC power.
The 2B emergency diesel generator (EDG), which is air-cooled and located away fromthe ocean in the common spent fuel pool building, did not flood and continued to operate.
The electrical switchgear for the 2B EDG, however, is located below grade in thebuilding, was wetted by the tsunami, and subsequently failed. Lighting and indications
were lost as AC and DC power systems failed, including all indications for HPCI and

21

INPO 11-005
RCIC. Control room lighting, including emergency lighting, failed completely, leaving
the control room dark. TEPCO management made an emergency declaration because of
the loss of all AC power and notified the government and associated authorities.
The operators were not sure if RCIC was still operating because the indicating lights hadgone out. High pressure coolant injection, which requires DC power to operate, became
unavailable when flooding inundated the DC distribution system. At 1636, anotheremergency was declared because of the inability to determine reactor water level and thestatus of injection into the core. With the possibility of no core cooling and extensive
damage to the site, workers began to investigate methods for venting containment
without power and reviewing methods for injecting water into the reactor using the fire
protection system or fire engines.
Batteries and cables were taken to the control room. After confirming the wiring usingdrawings, workers planned to connect batteries directly to instrument panels in thecontrol room. A top priority was to ascertain the status of water injection into the reactor
pressure vessel. Restoration work focused on connecting batteries to the reactor waterlevel indicator, which uses DC power.
Workers confirmed the submergence and damage of power panels (high-voltage
switchgear) by measuring the insulation resistance. They also confirmed two power
centers were not damaged and could be energized. They identified the possibility ofhigh-pressure water injection using the control rod drive system and standby liquid
control systems if portable generators could restore power.
At 1712 (T plus 2.4 hours), the site superintendent directed workers to investigate
methods for injecting water into the reactor using the fire protection system. Emergency
Response Center personnel discussed using alternative water injection, which was
implemented as an accident management measure. This included the use of fire
protection tanks and fire engines, which had been deployed after the Niigata-Chuetsu-oki
earthquake in 2007. The operators established an alternative water injection flow path
via the residual heat removal system by manually opening valves in the turbine andreactor building to establish an injection path after the reactor pressure decreased to 100
psig (0.69 MPa gauge). This pressure is sufficiently low to allow the fire pump to inject.
At 2049 (T plus 6.1 hours), workers restored some temporary control room lighting in theunits 1-2 control room when a small portable generator was installed. Criticalindications, such as reactor water level and the status of RCIC, remained out of service.
Without functional indications, operators reported that reactor water level may lower
below the top of active fuel (TAF). TEPCO informed the government that the estimatedtime for Unit 2 reactor water level to lower to TAF was 2140. Immediately following
this report, the Prime Minister expanded the evacuation zone to a 1.9 mile (3 km) radiusof the station and directed inhabitants within a 1.9 to 6.2 mile (3 to 10 km) radius to takeshelter.

22

INPO 11-005
At 2150 (T plus 7.1 hours), workers restored multiple indications, including reactor waterlevel indication, in the control room. Indicated reactor water level was 134 inches (3,400mm) above TAF. Both the restoration of indication and the indicated reactor water level
were reported to the government.
Workers next restored one channel of reactor pressure indication and containmentpressure indication. Unit 2 reactor pressure indicated 914 psig (6.3 MPa gauge) andcontainment pressure indicated 20 psia (0.14 MPa abs) at 2325 (T plus 8.7 hours).
Dose rates in the control room (shared with Unit 1) started increasing, most likely as aresult of ongoing core damage and releases from Unit 1. Operators continued to work on
restoring indications while lining up a fire engine to supply another injection source.
Just before midnight, the first of several power supply vehicles (mobile generators) began
to arrive on site. The generators were limited in their effectiveness because they could
not be connected to the station electrical distribution system as a result of the extensive
physical damage caused by the tsunami and flooding. The first mobile generator wasplaced adjacent to Unit 2, and workers began to lay temporary cables from the generator
to the associated distribution panel for the standby liquid control pumps.
Operators reviewed drawings to determine if they could open the valves required forventing. Based on their reviews of piping and instrumentation diagrams, accident
management procedures, and the venting procedure, operators developed a method forventing containment. They prepared a venting plan and began to review the locations of
the vent valves.
To confirm the operation of RCIC, operators were sent to locally inspect the system.
Wearing breathing protection and boots, workers attempted to verify the condition of
RCIC, but field conditions made this very difficult. Work that would normally takeapproximately 10 minutes required more than one hour to complete. The RCIC room
was dark, and the water level in the room came nearly to the top of the worker's boots, sothey turned back without getting close enough to check system operation. While theworkers could not get close to the system, they could hear faint metallic sounds, which
they interpreted as indicating the system was operating. Because of the lack ofcommunication methods, the workers had to return to the control room to report theirfindings.
At about 0200 on March 12, workers made another attempt to verify RCIC operation. On
this attempt, the workers discovered that the water level in the RCIC room had increased,
and they could not enter the room. They checked reactor pressure and RCIC pumpdischarge pressure on an instrument rack in the reactor building. The RCIC pumpdischarge pressure was high, so the workers concluded RCIC was operating. The
workers returned to the control room to report that RCIC was in operation on Unit 2.
With this information, opening the Unit 1 containment vent valves was made a priority.
Operators proceeded with the lineup to vent Unit 1 and, at the same time, monitored
parameters on Unit 2.

23

INPO 11-005

By 0500 (T plus 14.2 hours), workers were directed to wear full face masks with charcoal
filters and coveralls in both the control room and in the field. Increasing dose rates in
the Unit 1 control room caused operators to periodically move to the Unit 2 side of theroom. By this point, the level in the condensate storage tank was likely to be low and the
suppression chamber level high, so the RCIC water supply switched from the condensate
storage tank to the suppression chamber.
At 1530 (T plus 24.7 hours), the temporary cable connection between Unit 2 and thepower supply vehicle was completed. Operators were only moments away fromenergizing the Unit 2 standby liquid control system when, at 1536, an explosion occurredin the Unit 1 reactor building. Debris struck and damaged the cable and the power supply
vehicle, and the generator stopped. A review of satellite pictures revealed that the
explosion also caused a blowout panel in the Unit 2 reactor building to open, which
resulted in a loss of secondary containment integrity. Fieldwork was suspended andworkers evacuated to the Emergency Response Center (ERC). The next day, workers
attempted to start the power supply vehicle; however, the current surge relay actuated andthe generator failed.
At 1730 (T plus 26.7 hours), the site superintendent ordered preparations to proceed forventing the Unit 2 containment. Injection with RCIC continued and indicatedcontainment pressure remained relatively stable at 29–44 psia (200-300 kPa abs).
Concurrent preparations started for lining up to vent both the units 2 and 3 containments.
Considering the complications on Unit 1, operators planned to manually open the ventvalves while the dose in the field was low, aligning the suppression chamber vent to therupture disk.
The motor-operated containment vent valve (MO-271) was opened at 0810 on March 13(T plus 41.4 hours). At the time, containment pressure indicated approximately 50.8 psia
(0.35 MPa abs). At 0855, indicated containment pressure reached 52.9 psia (0.365 MPaabs), below the design pressure of 55.1 psig (0.38 MPa gauge), then began to lower. The
venting lineup was not yet complete. At 1015 (T plus 43.5 hours), the site superintendent
directed operators to vent the Unit 2 containment (see Figure 7.4-5). Workers used thesmall generator in the control room, which had been installed to restore some lighting, toenergize the solenoid for the large air-operated suppression chamber vent valve (AO205). At 1100 (T plus 44.2 hours), the vent lineup was completed but indicatedcontainment pressure was lower than the 62 psig (427 kPa gauge) pressure necessary to
open the rupture disk and allow venting, and the rupture disk remained intact. The site
superintendent then ordered workers to prepare for seawater injection into the reactor.
Reactor water level continued to be maintained by RCIC, but conditions were degrading
and operators monitored reactor water level for indications of a RCIC failure. At thedirection of the site superintendent, workers began staging hoses and equipment tosupport injecting into the Unit 2 reactor using fire engines when needed. Plans weremade to depressurize the reactor quickly when needed, allowing operators to switch corecooling to seawater via the fire protection system rapidly.

24

INPO 11-005

On March 14 at 1101 (T plus 68.3 hours), a hydrogen explosion occurred in the Unit 3
reactor building. The explosion damaged the temporary power supply used to open the
Unit 2 suppression chamber vent valve (AO-205), causing the valve to fail closed.
Indicated containment pressure was stable around 66.7 psia (460 kPa abs), just below the
rupture disk pressure. Debris damaged the fire engine and hoses that had been staged to
inject seawater into the reactor. All field work was stopped and workers evacuated to the
ERC for accountability.
In the ERC, workers became concerned about the ability to depressurize the reactor to
inject water with fire engines. The suppression chamber was saturated, indicating 301oF(149.3oC) and 70.5 psia (486 kPa abs). With the suppression chamber saturated, thesafety relief valves (SRVs) may not provide a quick reduction in reactor pressure needed
to switch core cooling to the fire protection system. Based on this concern, TEPCO
decided to vent Unit 2 containment first to remove energy, then open a safety relief valve(SRV) to reduce reactor pressure and switch from RCIC to seawater injection.
Work to prepare fire engines and hoses recommenced at 1305 (T plus 70.3 hours).
Because of scattered radioactive debris, workers shifted the seawater source to theshallow draft quay (harbor) instead of the main condenser backwash valve pit of Unit 3.
In the high radiation environment surrounded by scattered rubble, workers prepared a
new water injection line using available fire engines and hoses.
At 1318 (T plus 70.5 hours), reactor water level began to trend downward. By 1325,
operators concluded that RCIC had failed and core injection was lost. Attempts to restart
RCIC were unsuccessful. At the time of the trip, indicated reactor water level wasapproximately 95 inches (2,400 mm) above the top of active fuel (TAF) and containmentpressure was 67 psia (465 kPa abs). Calculations revealed that reactor water level would
reach the top of active fuel at approximately 1630. Workers continued to focus onventing containment, but ongoing aftershocks and evacuations delayed the work.
Because of delays in opening the suppression chamber vent valve (AO-205), TEPCO
changed the priority from venting containment to depressurizing the reactor. At
approximately 1600 (T plus 73.2 hours), an evacuation order was lifted, which allowed
field work to recommence; and the site superintendent directed that the efforts to ventcontainment be performed in parallel with reactor depressurization. Indicated reactor
water level had now decreased to 12 inches (300 mm) above TAF. By 1620, workers
realized that a temporary air compressor was not providing sufficient air pressure and thatthe large air-operated suppression chamber vent valve was not opening.
At 1630, indicated reactor water level had decreased below the top of active fuel, andoperators began work to open an SRV and align seawater injection into the reactor. Highradiation levels on site, caused by the radioactive debris from the units 1 and 3 reactor
building hydrogen explosions, precluded continuous manning of the fire engines.
Workers started a fire engine and lined up injection so water would flow when reactor
pressure was low enough. Periodic tours were conducted to check fire engine operation.

25

INPO 11-005

Because of a lack of power, temporary batteries were necessary to open the SRV.
Batteries were gathered from cars, carried to the control room, and connected. However,the voltage was insufficient, so additional batteries were scavenged and added. Operators
attempted to operate several SRVs without success. With no injection, reactor waterlevel decreased. The lack of core cooling likely resulted in core damage and thegeneration of hydrogen from the high-temperature interaction of steam and zirconium
inside the reactor.
Efforts to depressurize the reactor continued to about 1800 (T plus 75.2 hours), whenenough batteries had been installed and sufficient power existed to open the SRV. The
high suppression chamber temperature resulted in reactor pressure lowering more slowly
than desired. Reactor pressure indicated 1,015 psig (6.998 MPa gauge) at 1634 and 881
psig (6.075 MPa gauge) at 1803 and lowering; but the open SRV resulted in additional
inventory loss from the reactor, and indicated reactor water level continued to lower. At
1822 (T plus 75.6 hours), indicated reactor water level decreased to 146 inches (3,700mm) below TAFoffscale low, indicating the core may have been completely
uncovered. The open SRV resulted in an energy transfer into containment, and hydrogenand radioactive gases escaped from the reactor and accumulated inside containment.
However, indicated containment pressure remained relatively constant. It is postulated
that some of these gases entered the reactor building from leaks in the containment. The
open blowout panel in the reactor building prevented the hydrogen from reaching an
explosive level; however, it allowed the subsequent escape of radioactive materials to theenvironment.
Workers had still not been able to establish a containment vent path. TEPCO personnel
now suspected that the solenoid on the air supply valve had failed, preventing the large
air-operated suppression chamber vent valve from being opened. Workers attempted toopen the small air-operated suppression chamber vent valve to vent the containment.
Reactor pressure continued to lower slowly; and by 1903 (T plus 76.3 hours), reactor
pressure had reached 91 psig (0.63 MPa gauge), below the discharge pressure of the fireengine. The injection of seawater should have commenced; but at 1920, a worker touring
the area discovered the fire engine had run out of fuel. Workers added fuel and restarted
the engine at 1954, establishing seawater injection into the core. A second fire engine
was started and aligned to inject at 1957, but reactor water level remained below theindicating range.
At 2100 (T plus 78.2 hours), operators opened the small suppression chamber airoperated vent valve (AO-206), establishing the venting lineup (other than the rupturedisk). Indicated containment pressure remained slightly lower than the 62 psig (427 kPa
gauge) working pressure of the rupture disk, so venting did not occur. The vent valvesremained open, and operators monitored containment pressure.
Between 2037 and 2118, reactor pressure began to increase and exceeded the dischargepressure of the fire engine. At 2120, operators opened a second SRV to increase the rateof reactor depressurization. Forty minutes later (2200), indicated reactor water level

26

INPO 11-005
recovered to 63 inches (1,600 mm) below TAF. However, containment pressure wasnow increasing. Near the site boundary, dose rates increased to as high as 76 mrem/hr(760 µSv/hr).
At 2240, another unexpected increase in reactor pressure occurred. Reactor pressure
indications increased from 62 psig (0.428MPa gauge) at 2240 to 264.4 psig (1.823MPa
gauge) at 2250. At the same times, indicated reactor water level lowered from 27.6
inches (700 mm) below TAF to 63 inches (1,600 mm) below TAF. The cause of theincrease in reactor pressure has not been identified. With reactor pressure above theshutoff head for the fire engines, reactor water level again began to decrease, exposing
more of the fuel. At approximately 2330, indicated reactor pressure again began to trend
down, but reactor water level indication remained off-scale low. As the reactor
depressurized, containment pressure increased, indicating a peak value of 108.8 psia
(0.75 MPa abs) at 2344.
Operators began to recognize some abnormalities in their indications. Containmentpressure was well above the rupture disk pressurebut the rupture disk had not failed.
Additionally, indicated drywell pressure was trending upward and had increased above
102 psia (0.7 MPa abs), whereas indicated suppression chamber pressure was stable atabout 43.5 to 58 psia (300-400 kPa abs), below the rupture disk pressure. The nonunified pressures indicated a problem. As indicated suppression chamber pressure waslower than the working pressure of the rupture disk and indicated drywell pressure
increased above the design pressure, the operators decided to open the small air-operated
drywell vent valve (AO-208) to vent directly from the drywell to reduce pressure.
Two minutes after midnight on March 15, the operators opened the small air-operated
drywell vent valve (AO-208). The vent line lineup was complete, except for the rupture
disk that remained closed. Containment pressure remained stable at approximately 109
psia (750 kPa abs). The operators rechecked their lineup and found that the small airoperated drywell vent valve had already failed closed. They continued to work toward
establishing a containment vent path for Unit 2; but at about 0600 (T plus 87.2 hours), a
loud noise was heard in the area around the torus and suppression chamber pressure
indication failed low. At approximately the same time, a hydrogen explosion occurred in
the Unit 4 reactor building. The loud noise in Unit 2 was widely reported as anotherexplosion, which was accompanied by torus pressure lowering to atmosphericpressurebut this was later determined to be incorrect. A review of the seismic
instrumentation revealed that, if something did fail in Unit 2, the failure did not generate
the same shock wave or force as an explosion. Additionally, the suppression chamberpressure reading 0.0 psia (0.0 MPa abs) is an indication of a failed instrument, not anindication of atmospheric pressure. Indicated drywell pressure remained stable atapproximately 106 psia (0.73 MPa abs) and reactor water level indicated 110 inches(2,800 mm) below TAF. The Unit 2 containment was not vented, and the cause for thecontainment pressure changes has not been determined.
Following the explosion in the Unit 4 reactor building, dose rates at the main gateincreased to 58.4 mrem/hr (583.7 µSv/hr). All nonessential personnel (approximately

27

INPO 11-005
650 people) were evacuated to Fukushima Daini (approximately 4.3 miles, (7 km) away),leaving 70 people at the station. Shift operators periodically went to the control room tolog containment pressure and other critical parameters, but no values were recorded from0720 until 1125. When containment pressure was checked at 1125, it indicated 22.5 psia
(0.155 MPa abs). With no reports of steam being discharged from the Unit 2 vent stackand no changes to the venting lineup, TEPCO suspects the loud noise, instrument failure,and subsequent containment depressurization to be indicative of a potential breach of thecontainment.
As the day continued, the operators lined up a fire engine to inject seawater into thereactor through the core spray system and commenced injecting seawater at 1954 on
March 14. Injection was occasionally challenged as reactor pressure varied. Boron wasadded to the water source to address criticality concerns.
This situation continued over the next several days as site personnel attempted to restore
electrical power to the unit. Off-site power was restored to Unit 2 on March 20, ninedays after the event.
4.3

Unit 3 Narrative

On March 11, 2011 at 1446 (T=0), an earthquake caused a loss of off-site power and an
automatic reactor scram. All control rods inserted and several actions occurred, including
a loss of feedwater and condensate and main steam isolation valve closures, as expected
because of the loss of off-site AC power. The emergency diesel generators started andloaded in response to the loss of off-site power and supplied power to the safety systems.
After waiting for the shaking from the earthquake to stop, the operators manually tripped
the main turbine because of high vibrations and subsequently broke main condenser
vacuum to help stop the main turbine.
After the scram, reactor water level initially dropped as expected because of thecollapsing steam voids. At 1505, operators initiated RCIC to maintain reactor pressure
and water level after the scram. Twenty minutes later, RCIC automatically shut downbecause of a high reactor water level. With reactor water level high in the control band,the operators monitored reactor water level and waited for the level to lower before
restarting RCIC.
At 1527 (T plus 41 minutes), the first of a series of seven tsunamis, generated by theearthquake, arrived at the station. The second tsunami, which arrived at 1535, flooded
and damaged the intake structure. By 1538 (T plus 52 minutes), the tsunami had begun
to cause flooding in the turbine building basement. The flooding wetted or submerged
the Unit 3A and 3B emergency diesel generators and the electrical distribution systems,
resulting in a gradual loss of all AC and most DC power. Lighting and indications werelost as AC and DC power systems failed. Normal control room lighting failed
completely, but some DC power remained for emergency lighting and indications.
TEPCO management made an emergency declaration because of the loss of all AC power
and notified the government and associated authorities. Two field operators were noted

28

INPO 11-005
as missing from the units 3 and 4 operating crew. The operators were later found to havedrowned after being trapped in the Unit 4 turbine building basement when the tsunami
flooded the building.
While the tsunami caused the loss of some DC power systems, including some of thecontrol board instrumentation and controls, the operators had indication of reactor
pressure and reactor water level. Both HPCI and RCIC remained available for injection.
Operators used safety relief valves as needed to control reactor pressure.
At 1603 (T plus 1.3 hours), indicated reactor water level had lowered and RCIC was
restarted to restore reactor water level. With RCIC in service, reactor water level wasmaintained approximately 157 inches (4,000 mm) above TAF by narrow range
indication. At 2158, a small portable generator was used to restore lighting in the units
3-4 control room. These conditions were maintained through the evening.
The next day (March 12) at 1136, RCIC shut down unexpectedly and could not berestarted. At this time, no water was being injected into the reactor. Operators requested
that a fire engine be dispatched to prepare for injecting water into Unit 3, but all fireengines were being used to mitigate the ongoing accident in Unit 1. Requests for off-site
fire engines were initially unsuccessful because the roads were impassable. Primarycontainment pressure indication slowly increased, peaking at 57 psia (0.39 MPa abs),
below the design pressure of 55 psig (0.38 MPa gauge).
At 1235 (T plus 21.8 hours ), an hour after RCIC tripped, HPCI automatically started ona low-low reactor water level signal and began to restore reactor water level. Operators
throttled open full-flow test valves to return some of the HPCI flow to the suppression
chamber, possibly in an attempt to prevent a high-level trip. By 1635, indicated reactor
water level had recovered to 15.7 inches (400 mm) by wide range indication, or
approximately 180 inches (4,570 mm) above TAF. The HPCI system was drawing aconsiderable amount of steam off the reactor, and reactor pressure had begun to decreasesignificantly. At 1700 (T plus 26.2 hours), reactor pressure indicated 421 psig (2.9 MPa
gauge) and lowering.
The site superintendent ordered preparations to proceed for venting the containment.
Knowing that venting would be required eventually, operators commenced preparations
for lining up to vent the Unit 2 and Unit 3 containments. Operators planned to manually
open the vent valves while the dose rates in the reactor building were currently low.
As the evening continued, station batteries depleted, and DC-powered instruments began
to degrade and fail. At 2027, the accident management indications for drywell pressure,suppression chamber pressure, and suppression chamber water level failed. Nine minutes
later, at 2036, reactor water level indication was lost. The last indicated reactor waterlevel before the failure was 53 inches (1,350 mm) by wide range instrument (217 inches
or 5,520 mm above TAF) , but only 16 inches (400 mm) above TAF by fuel zone
instruments.

29

INPO 11-005
After 2100 (T plus 30.2 hours), operators started a review of the vent procedures to
identify the sequence and location of vent valves, which were written on a whiteboard.
Emergency response workers reviewed the vent procedure for Unit 1. Based on thereview and the accident management procedure for Unit 3, they developed a plan to
manually vent Unit 3 and provided the plan to the control room operators.
At 0242 on March 13 (T plus 35.9 hours), the HPCI system tripped. At the time of thetrip, DC power was failing and reactor pressure was low, indicating approximately 84
psig (0.58 MPa gauge). Operators attempted to restart HPCI but were unsuccessful
because the batteries were exhausted. Operators were unable to restart RCIC locally.
They next attempted water injection by diesel-driven fire pump, but reactor pressure wastoo high. With HPCI isolated and with no safety relief valves (SRVs) available becauseof the loss of DC, reactor pressure quickly increased well above the discharge pressure of
the fire pump, preventing water injection.
Injection into the reactor was lost, and at approximately 0415 on March 13 the reactor
core started to uncover. As the core uncovered, core damage commenced, and the hightemperature interaction of steam and zirconium began, generating large amounts ofhydrogen in the reactor.
The operators understood they needed to depressurize the reactor but had no method of
opening an SRV. All of the available batteries had already been used, so workers weresent to scavenge batteries from cars and bring them to the control room in an attempt toopen an SRV.
At 0450 (T plus 38.1 hours), workers attempted to open the large air-operated
suppression chamber containment vent valve (AO-205). To open the valve, workers used
the small generator to provide power to the valve solenoid. An operator checked thevalve indication locally in the torus room, but the valve indicated closed. The torus room
was very hot because of the previous use of RCIC, HPCI, and SRVs; and the room wascompletely dark, which made a difficult working environment. By 0500, reactor pressure
had exceeded 1,070 psig (7.38 MPa gauge), reactor water level indicated 79 inches(2,000 mm) below TAF and lowering, and containment pressure indicated 52.2 psia (0.36
MPa abs).
At 0515 (T plus 38.5 hours), the site superintendent instructed operators to complete thelineup for venting the containment (see Figure 7.4-5). The operators energized thesolenoid for the large air-operated suppression chamber vent valve (AO-205), but thevent valve remained closed. Operators determined that there was insufficient air pressure
to operate that valve. The operators replaced the temporary air bottle, and the valve
opened. The motor-operated vent valve, however, had not yet been opened, so the ventlineup was not complete.
With no ability to inject into the reactor, and containment pressure indicating 66.7 psia
(0.46 MPa abs) and increasing, operators aligned a fire engine to the containment spraysystem in an attempt to reduce containment pressure. Containment pressure, however,30

INPO 11-005
continued to increase. While not confirmed, the high containment temperature andpressure may have caused the drywell head seal and containment penetrations to degrade
and begin to leak. By 0735 (T plus 40.8 hours), indicated reactor water level had loweredto the bottom of the fuel zone indication, indicating the core may have been completely
uncovered.
At 0835 (T plus 41.8 hours), operators opened the motor-operated containment ventvalve. At 0841, they opened the large air-operated suppression chamber vent valve,
completing the vent lineup except for the rupture disk. A short time later, steam wasobserved discharging from the vent stack and dose rates at the site boundary increased to88.2 mrem/hr (882 µSv/h). The containment vent rupture disk had opened, and pressure
began to decrease. The maximum indicated containment pressure was 92.4 psia
(0.637 MPa abs) at 0910.
At approximately the same time, workers had scavenged enough batteries to power theSRVs, and the operators opened an SRV to depressurize the reactor. By 0925, reactor
pressure had decreased sufficiently to start borated fresh water injection. Reactor waterlevel recovered and increased above the top of active fuel.
By 1030 (T plus 43.7 hours), knowing that the site was low on fresh water, the site
superintendent ordered workers to be ready to commence injection using seawater.
Workers staged equipment in preparation for a quick transition to seawater injection.
They looked for seawater sources, including accumulated water in the basement of theUnit 4 turbine building. Workers broke through the truck bay doors and attempted tomove a fire engine into place, but this plan was unsuccessful. They also consideredtaking water from the discharge channel of Unit 4 or the training center pool, but this also
did not work. Workers elected to use a Unit 3 circulating water reversing valve pit,
which had been flooded by the tsunami, as the water source for the fire engines.
A short time later, at 1117, the air-operated suppression chamber vent valve (AO-205)
was found closed. The air cylinder being used to provide air was depleted because ofleakage. The workers attempted to lock open the valve locally, but they were notsuccessful because of the adverse conditions in the torus room. The room was dark andhot, and the torus was shaking because of the open SRV. Workers eventually replaced
the air bottle, and the air-operated valve was reopened. Similar problems challenged thecontainment vent lineup over the next few days. An engine-driven air compressor waslater installed to resolve these issues.
The fresh water supply was depleted at 1220 (T plus 45.6 hours), and injection into thereactor stopped. Workers had prepared to make a swift change to seawater injection, but
an aftershock and subsequent evacuation order occurred while the work was being carriedout, so the swap was delayed. By 1300, indicated reactor pressure had decreased to 28
psig (0.19 MPa gauge), and reactor water level indicated approximately 79 inches (2,000mm) below TAF by the lowest indication.

31

INPO 11-005
Seawater injection was established at 1312. However, indicated reactor water levelremained below the top of active fuel. The dose rates measured at the air lock to Unit 3
increased to 30 rem/hr (300 mSv/hr). By 1530, indicated reactor water level was
74.8 inches (1,900 mm) below TAF. Dose rates in the Unit 3 side of the control room
reached 1.2 rem/hr (12 mSv/hr), and operators moved to the Unit 4 side of the room in an
attempt to minimize their exposure.
At 0110 on March 14 (T plus 58.4 hours), injection was stopped when the water level inthe reversing valve pit became low. Workers began refilling the pit using other fireengines. Reactor water level indicated 88.6 inches (2,250 mm) below TAF.
At 0200 (T plus 59.2 hours), operators noticed that containment pressure was trending
upward, indicating at 38.4 psia (0.265 MPa abs) and increasing. Because of ongoingproblems with the large air-operated suppression chamber vent valve (AO-205), workers
decided to open the small air-operated suppression chamber vent valve (AO-206).
Containment pressure continued to increase. The workers could not maintain the valve
open because of a loss of air pressure and a loss of DC power to the solenoid.
To restore injection into the reactor, workers moved the fire engine around, allowing thesuction hose to drop deeper into the valve pit. At 0320 (T plus 60.6 hours), seawater
injection into the reactor was restored. The injection rate was not sufficient, andindicated reactor water level continued to lower. By 0430, reactor water level was below
the indicating range of the fuel zone instrument, indicating the core may have been
completely uncovered.
Additional fire engines, which had been requested previously, began to arrive aroundsunrise. The fire engines were delayed because the drivers needed to be changed tostation workers off site because of the contamination and radiation levels at the site.
Workers began using two fire engines that had arrived from off site to pump seawaterfrom the shallow-draft quay (harbor) into the reversing valve pit. At 1053 (T plus 68.1
hours), seven 5-ton Self-Defense Force water supply vehicles arrived and began to refill
the reversing valve pit. However, continued problems with the containment vent lineup
had resulted in indicated containment pressure increasing and stabilizing at
approximately 75.4 psia (0.52 MPa abs).
A large hydrogen explosion occurred in the Unit 3 reactor building at 1101 on March 14.
The explosion destroyed the secondary containment and injured 11 workers. The large
amount of flying debris from the explosion damaged multiple portable generators and thetemporary power supply cables. Damage to the fire engines and hoses from the debris
resulted in a loss of seawater injection. Debris on the ground near the unit was
extremely radioactive, preventing further use of the main condenser backwash valve pit
as a source of water. With the exception of the control room operators, all work stopped
and workers evacuated to the Emergency Response Center for accountability.

32

INPO 11-005
An undamaged fire engine was moved to the shallow-draft quay, and at 1630 a new
injection lineup was completed. Fire engines and hoses were rearranged to injectseawater directly from the quay into both units 2 and 3 reactors.
This situation continued over the next several days as site personnel attempted to restore
electrical power to the unit. Off-site power was restored to Unit 3 on March 22, 11 days
after the event.
4.4

Unit 4 Narrative

Unit 4 was shut down and had been in a refueling outage since November 30, 2010. Allof the fuel had been removed from the reactor and placed in the spent fuel pool tofacilitate shroud work. The reactor was disassembled with the head off at the time of theearthquake. The cavity gates were installed, isolating the spent fuel pool from the upper
pools. Spent fuel pool temperature was approximately 80.6oF (27oC).
Following the earthquake and tsunami, the operators in the units 3-4 control room
focused the majority of their efforts on stabilizing Unit 3. Because of the low decay heatload in the Unit 4 spent fuel pool, operators may not have been immediately concernedwith taking action to fill or cool the spent fuel pool.
On March 15, however, a hydrogen explosion occurred in the Unit 4 reactor building.
This was unexpected, as workers did not believe there was enough decay heat in the pool
to result in overheating and the subsequent high-temperature interaction of zirconium andwater to produce hydrogen gas. After the explosion, however, some people suspectedthat the spent fuel was overheating, and they became concerned about subsequenthydrogen explosions caused by the fuel in the spent fuel pools on other units. Some
resources may have been diverted from attempts to stabilize the reactors, as extensive
actionssuch as helicopter water drops and the use of water cannonswere taken in an
attempt to refill the pools. Subsequent analysis and inspections determined that the waterlevel in the Unit 4 spent fuel pool never dropped below the top of fuel and that no
significant fuel damage had occurred.
There are various theories regarding the cause of the hydrogen explosion in Unit 4.
Hydrogen water chemistry was used at Fukushima Daiichi, and hydrogen was also used
to cool the main generators. Both of these sources of hydrogen were reviewed, and it
appears that neither source caused the Unit 4 explosion.
The most widely accepted theory is associated with the backflow of gases from Unit 3
during venting. The containment vent exhaust piping from Unit 3 is connected to theUnit 4 exhaust piping. The dampers on the Unit 4 standby gas treatment system (SGTS)
are air-operated and fail open on a loss of power or air (except the cross-connect between
SGTS filter trains). Additionally, the system does not have a backflow damper installedin the piping that connects to Unit 3. With no power or air, and no fans in service todirect the gases from Unit 3 up the exhaust stack, the exhaust gases from Unit 3 would bedirectly aligned to the Unit 4 SGTS filters. This piping arrangement may have allowed33

INPO 11-005
gases from the Unit 3 containment to be vented to the Unit 4 reactor building via reverse
flow through the Unit 4 standby gas treatment system.

Figure 4.4-1 Unit 4 Standby Gas Treatment System Hydrogen Flow Path
To confirm this theory, on August 25 TEPCO personnel performed radiation surveys on
the Unit 4 SGTS filters. The survey results, shown in Figure 4.4-2, revealed higher
radiation levels at the locations closer to Unit 3. These survey results support theconclusion that there was backflow from Unit 3 to Unit 4. Further inspections are needed
to confirm this theory; but based on this information, it is postulated that the hydrogen
explosion in the Unit 4 reactor building was caused by hydrogen from Unit 3.

34

INPO 11-005

Figure 4.4-2 Unit 4 Standby Gas Treatment System Filter Survey Results
4.5

Spent Fuel Pools and Dry Cask Storage
1

2

3

4

5

6

In the reactor

400

548

548

0

548

764

Spent fuel assemblies
in the spent fuel pool

292

587

514

1,331

946

876

New fuel assemblies
100
28
52
204
in the spent fuel pool
Water volume (ft3)
36,021 50,323 50,323 50,323
Table 4.5-1 Spent Fuel Pool Data

48

64

50,32352,866

Unit
Number of fuel assemblies

BackgroundAs shown, the spent fuel pools (SFPs) for units 14 contained different amounts of spent
fuel at the time of the event, generating different heat loads. The Unit 4 SFP had thegreatest heat load because the entire core had been offloaded into the SFP to support
ongoing outage work.
Approximately 60 percent of the spent fuel on site is stored in a separate building in acommon spent fuel pool. This pool contained 6,375 fuel assemblies (about 80 percent ofpool capacity), but the heat load was very low because the assemblies were stored in theirrespective units’ SFPs for 19 months or longer before being set in the common pool.
Calculations determined that cooling can be lost to this pool for 30 days before it
35

INPO 11-005
becomes a concern. The common spent fuel pool uses fans and air for cooling, so
cooling is maintained if the seawater ultimate heat sink is lost; however, AC power isrequired to power the fan motors and circulating pumps.
Dry cask storage is also used for spent fuel. At the time of the event, the station had ninecasks containing 408 spent fuel assemblies.
There are no mixed-oxide fuel assemblies in any of the spent fuel pools or in the dry cask
storage facility.
Event Progression
Spent fuel pool cooling flow was lost for all pools because of the loss of off-site power
and was not restored after the emergency diesel generators started. The existing water
inventory in the spent fuel pools provided sufficient cooling to remove decay heat, aslong as the fuel remained covered. Unconfirmed reports were that sloshing of the waterin the SFPs during the earthquake resulted in a loss of some inventory; however, this has
not been verified. After the tsunami impacted the site, operators were able to use the 6B
emergency diesel generator (EDG) to provide power to cooling systems for the Unit 6
spent fuel pool. After installing temporary cables, the 6B EDG provided power to Unit 5
spent fuel pool cooling.
Hydrogen explosions in the units 1, 3, and 4 reactor buildings, coupled with the loss ofthe blowout panel in Unit 2, resulted in the SFPs of all units being exposed toatmosphere. The explosions may have also caused additional inventory to be lost fromthe pools. After the explosion on Unit 4, a concern arose that the SFP may have boiled
dry and resulted in fuel damage and hydrogen generation. In response to this concern,
station personnel took numerous actions to ensure the spent fuel had sufficient cooling.
Fire engines from the surrounding area, including water cannons from the Japan SelfDefense Force, were brought to the site to spray water into the SFPs to keep the fuelassemblies covered. The water cannons could only reach the floor surrounding the SFP,
so little water got into the pool. On March 17, helicopters attempted to drop large loads
of water into the SFPs. Because of high radiation levels, the helicopters needed tomaintain a high altitude; and that, coupled with the prevailing winds, resulted in thiseffort not being successful.
Concrete pumping trucks were flown to Japan for use in pumping water into the SFPs.
These trucks, which have long, articulated booms normally used for transferring concrete,
were lined up to pump water into the spent fuel pools on March 18. This operation was
successful; however, the trucks did not start refilling the SFPs until several days after the
event began. The delay in refilling the SFPs may have contributed to increased radiationlevels in the area around the spent fuel pools because less shielding was provided withthe reduced water level. The exact impact, however, was not known because radiationlevels were not monitored as the pools were refilled.

36

INPO 11-005
Subsequent analyses and inspections determined that the spent fuel pool water levels
never dropped below the top of fuel in any spent fuel pool and that no significant fueldamage had occurred. Current investigation results indicate that any potential fueldamage may have been caused by debris from the reactor building explosions.
An inspection of the dry cask storage facility revealed that, while the building wasdamaged by the tsunami, the dry storage casks do not appear to be damaged. The casks
were wetted by the tsunami, but they were not moved from their storage locations by theforce of the waves or debris. The fuel stored inside the dry casks has not yet beeninspected.

37

INPO 11-005
5.0Radiological Overview
Capabilities for monitoring radiological conditions effectively both on site and off sitewere severely hampered by the effects of the earthquakes and tsunamis. The earthquake
did not damage on-site monitoring systems, but few remained following the tsunamis.
For example, most electronic personnel dosimeters, computer systems for activating andrecording dose from these devices, installed contamination monitors, and many portable
survey instruments were lost to the flooding. Installed radiation monitors essential formonitoring core, containment, and spent fuel pool conditions were lost when the tsunamis
wetted the electrical distribution equipment. In addition, radiological effluent,environmental, and meteorological monitors were lost.
During the event, Fukushima Prefecture officials directed several evacuations. Evacuees
were screened for contamination upon reporting to shelters. The screening criterion forevacuees reporting to shelters was originally 100,000 counts per minute (cpm). The
screening criterion was revised on March 20 to 0.1 mrem/hr at 4 inches (10 cm) to alignwith IAEA standards. Between March and June, 195,354 people were screened, with 102
persons being decontaminated below the limits.
Following the event, Japanese government officials estimated the amount of radioactivity
discharged into the air between March 11 and April 5 was equivalent to 1.7 E7 curies (6.3
E17 Bq).
On April 2, very high concentrations of radioactivity were identified in the harbor of thestation. The source was water accumulating in the turbine building, flowing through atrench, and leaking into the harbor. The magnitude of this release was estimated at 1.3
E5 curies (4.81 E15 Bq). This source was stopped after a couple of days. The ocean
around the plant was sampled daily following this discovery, and the maximum
concentration in late July was approximately 1 E3 pCi/l (3.7 E1 Bq/l) of cesium-137.
5.1

On-Site Capabilities

The loss of personnel monitoring capabilities resulted in the need for initial emergency
responders to share dosimeters, with only one worker on a team wearing dosimetry formany missions. The limited number of electronic dosimeters made measuring worker
dose difficult because individual passive dosimeters, such as thermoluminescentdosimeters, are normally not worn at the station.
With no process radiation monitoring indications available, operators were not aware ofchanges in radiological conditions until personnel in the field noted them. For example,
when returning the Unit 1 isolation condenser to service at 2130 on March 11, dose rates
in the reactor building increased more than expected. This is an indication that fueldamage may have already begun and likely created the first radiological release to theenvironment. However, operators were not aware of this condition until workers in theplant reported increases in their dose and elevated dose rates were measured outside thereactor building personnel air lock. Later on March 12, operators were unsure if attempts

38

INPO 11-005
to vent Unit 1 were successful because the radiological monitor on the plant vent stack
was inoperable.
The loss of installed radiological monitoring equipment and meteorologicalinstrumentation also contributed to TEPCO and off-site agencies relying on postulated
core damage scenarios to perform off-site dose projections.
5.2

Site Boundary and Off-Site Capabilities

Radiological monitors installed off site and at the site boundary (approximately 0.62
miles (1 km) from the plant) were also lost from either the widespread loss of power that
resulted from the earthquake or from the tsunamis. Twenty-three of 24 off-sitemonitoring posts, the off-site central monitoring facility, and all 8 site boundary
monitoring posts were out of service. Additionally, local government personnel who
were responsible for collecting off-site radiological data either could not be locatedfollowing the earthquake and tsunami or were supporting other disaster duties. This
resulted in no organized off-site radiological monitoring until March 16, when Ministryof Education, Culture, Sports, Science and Technology personnel, the ministry
responsible for environmental radiation monitoring, assumed this responsibility.
To compensate for the loss of installed monitors at the site boundary, station personnel
had a vehicle equipped with radiological and meteorological instruments. Monitoring
began at the site boundary on the inland side of the plant at 1700 on March 11. This
monitoring was normally limited to a single location. Although valuable data was
obtained, air samples were not taken; and, as the event progressed, the data was often notcollected in close proximity to the plume.
5.3

On-Site Dose Rates and Protective Measures

After off-site evacuations had been reported as complete at 0903 on March 12, operators
entered the Unit 1 reactor building in an attempt to vent containment. The reactor
building had elevated dose rates and a steam-filled environment. The operators wore
fire-fighting turnout gear and self-contained breathing apparatus and were provided withelectronic dosimeters set to alarm at 10,000 mrem (100 mSv) and potassium iodide forthyroid protection. The first team opened a valve on the second floor of the reactor
building, but elevated dose rates in the torus room required the second team to turn backbefore completing its mission. One of the operators received a dose of 10,630 mrem
(106.30 mSv) while in the torus room. This was reported to the authorities as anoverexposure because it was above the 10,000 mrem emergency exposure limit.
The explosion on Unit 1 at 1536 on March 12 spread highly radioactive debris around thesite. The door to the units 1-2 control room, which had been opened to allow workers to
run temporary power cables, was damaged by the explosion, allowing radioactive
material into the control room. The elevated dose rates and high levels of surface andairborne radioactivity around the site hampered efforts to stabilize units 2 and 3 andresulted in substantial dose to site workers. Operators in the units 3-4 control room were39

INPO 11-005
directed to wear charcoal respirators after the Unit 1 explosion. However, tocontinuously staff their posts, they needed to remove their respiratory protection on
occasion, such as to eat. In addition, a few of the operators wore conventional eyeglasses(spectacles kits were not a requirement), which prevented their respirators from sealing
properly. As a result, two operators in the units 3-4 control room received doses inexcess of the initial 10,000 mrem (100 mSv) and later 25,000 mrem (250 mSv)
emergency dose limit. One received 67,808 mrem (678.08 mSv), of which 59,000 mrem
(590 mSv) was internal. The other received 64,307 mrem (643.07 mSv), with a totalinternal dose of 54,000 mrem (540 mSv).
On March 13 at 1300, pressure in the Unit 3 drywell and torus began to decrease rapidly,
indicating a release from the containment. On-site dose rates as high as 30,000 mrem/hr(300 mSv/hr) were measured outside the Unit 3 personnel air lock. Operators in the units
3-4 control room had to move to the Unit 4 side because dose rates on the Unit 3 side
reached 1,200 mrem/hr (12 mSv/hr). The TEPCO medical chief directed site personnel
under 40 years of age to take potassium iodide, while older workers were given theoption.
On March 14, at 0700, Unit 3 containment was vented. Indication on a Unit 3 drywellradiation monitor had been recovered briefly just before the venting, and a dose rate of
16,700 rem/hr (167 Sv/hr) was recorded. The Unit 3 explosion severely impacted workwithin the plant. Debris on the ground had dose rates exceeding 1,000 rem/hr (10 Sv/hr),and workers establishing water injection for Unit 2 could no longer continuously monitor
fire engines because of the elevated dose rates.
On March 15, elevated dose rates around the site led TEPCO officials to evacuate
nonessential personnel at the site to Fukushima Daini. Seventy workers remained on site
and 650 evacuated to Fukushima Daini.
The high surface and airborne contamination levels around the site also resulted in the
Emergency Response Center becoming contaminated early in the accident. The buildup
of contamination was not recognized until radiological surveys in the building were first
performed on March 24. As a result, controls were not in place to prevent uptakes, andsome workers received substantial internal doses. For example, a female worker in thebuilding exceeded the 500 mrem (5 mSv) quarterly dose limit when she was determined
to have received 1,350 mrem (13.50 mSv) of internal dose.
On March 22 and 23, surveys of the airborne radioactivity and dose rates around the sitebegan to be collected and documented. The dose rates are documented on Figure 5.3-1.
Localized dose rates as high as 1,000 rem/hr (10 Sv/hr) were later discovered.

40

INPO 11-005

Figure 5.3-1 Site Dose Rates
Dose rates inside the reactor and turbine buildings were taken in April and May. Dose
rates 3.3 feet (1 m) above water that had accumulated in the turbine buildings ranged
from 0.3 rem/hr (0.003 Sv/hr) in Unit 4 to 100 rem/hr (1 Sv/hr) in Unit 2. The generalarea dose rates in the reactor buildings ranged from slightly below 1 rem/hr (0.01 Sv/hr)
to approximately 43 rem/hr (0.43 Sv/hr) in Unit 2. A localized area around a steam leakon Unit 1 had dose rates as high as 400 rem/hr (4 Sv/hr). Air samples collected in thereactor buildings measured iodine-131 concentrations equivalent to as high as 3 rem/hr(0.03 Sv/hr) of thyroid dose if breathed by an unprotected worker. Although highconcentrations of strontium isotopes, a strong beta emitter, have been discovered off site,measurements of beta dose rates have not been documented in available plant surveys.
As a result of the high beta contamination levels, two workers received beta burns when
they stepped in turbine building water.
5.4

Site Boundary and Off-Site Dose Rates and Protective Measures

Evacuations in the area surrounding the plant were first ordered at 2050 on March 11 forthe people living within 1.2 miles (2 km) of the site. The evacuation was extended to a
1.9 mile (3 km) radius at 2123, and those within 6.2 mile (10 km) were directed to shelter
within their homes.
The first increase in dose rates at the site boundary was detected at 0404 on March 12.
Dose rates near the main gate increased from the normal background of 0.007 mrem/hr(0.07 Sv/hr) to 0.1 mrem/hr (1.0 Sv/hr). This data, along with a decrease in drywellpressure, prompted Unit 1 control room personnel to conclude that a release had
occurred. At 0544, the Prime Minister extended the evacuation radius to 6.2 miles (10km), and a TEPCO press release at 0600 reported elevated dose rates around the site.
At 1020 on March 12, while Unit 1 operators were attempting to open the containment
vent valve, dose rates at the site boundary briefly reached as high as 38.5 mrem/hr (0.385
mSv/hr). Dose rates later spiked as high as 101.5 mrem/hr (1.015 mSv/hr) at the siteboundary, and steam was seen leaving the plant stack after operators vented Unit 1 torus

41

INPO 11-005
using a temporary air supply at 1410. Dose rates at the site boundary remained elevatedand had no discernable increase after the explosion that occurred on Unit 1 at 1536. The
evacuation radius was extended to 12.4 miles (20 km) at 1825. Over the next day, dose
rates remained elevated well above normal background with periodic increases, whichindicated that radiological releases continued to occur.
On March 13 at 1300, pressure in the Unit 3 drywell and torus began to decrease rapidly,
indicating a release from the containment. Dose rates at the site boundary increased to155.7 mrem/hr (1.557 mSv/hr).
There were no appreciable changes in site boundary dose rates on March 14 following
venting of the Unit 3 containment because the wind direction was blowing the plumetoward the ocean. However, the release was likely very large given that a Unit 3 drywellradiation monitor recovered briefly just before the venting indicated a dose rate of 16,700
rem/hr (167 Sv/hr). The wind continued to blow toward the ocean, and site boundary
dose rates did not increase when a hydrogen explosion occurred on Unit 3 at 1101. Dose
rates at the site boundary did increase to as high as 313 mrem/hr (3.13 mSv/hr) later inthe evening, and elevated dose rates were measured 62 miles (100 km) south of the plant.
On March 15 at 0820, site boundary dose rates began to increase rapidly to 821 mrem/hr(8.21 mSv/hr). This increase occurred approximately two hours after the pressure in the
Unit 2 drywell began to decrease. This release is likely responsible for much of theelevated dose rates later discovered off site because of deposition that occurred from therainfall in many areas during this release. Government officials directed that inhabitants
between 12.4 miles (20 km) and 18.6 miles (30 km) of the site remain sheltered. The
highest dose rates recorded at the site boundary were measured on March 16, when dose
rates reached 1,085 mrem/hr (10.85 mSv/hr).
Routine dose rate surveys began to be collected in the area outside the 12.4 mile (20 km)
radius of the plant on March 16. The highest dose rate of 17 mrem/hr (0.17 mSv) was
measured 19 miles (30.5 km) northwest of the station. The first air samples from the siteboundary, on March 22 and 23, had iodine-131 concentrations that were equivalent toapproximately 80 mrem (0.8 mSv) of thyroid dose each hour if inhaled by an unprotectedindividual. The concentration remained between 25 and 200 percent of this value until
April 18, 2011. Surveys inside 12.4 miles (20 km) began on March 30. The map shownin Figure 5.4-1, which was created based on survey results, is representative of these
initial surveys.

42

INPO 11-005

Figure 5.4-1 Initial Off-Site Survey Results
Government officials issued the first directive for the public to take potassium iodide andrestrictions on consumption of food and water on March 21. The directive for the
mandatory issuance of potassium iodide included the inhabitants of several villages andtowns within the affected area. Enough potassium iodide for 900,000 people wasdistributed within a 31-mile (50 km) radius of the plant. Because the evacuations had
already been completed, however, the potassium iodide was not issued to the population.
Food restrictions included the distribution of such items as spinach and raw milk from
Fukushima and a few nearby prefectures. While food restrictions have continued to be
added and lifted since the event, all water restrictions that had gone into effect were liftedby April 1, 2011, except in one small village where they remained until May 10, 2011 forinfants.

43

INPO 11-005
6.0

References
1. Report of the Japanese Government to the IAEA Ministerial Conference on Nuclear
Safety – The accident at TEPCO’s Fukushima Nuclear Power Stations, June 2011.
2. IAEA International Fact-Finding Expert Mission of the Fukushima Dai-Ichi NPP
Accident following the Great East Japan Earthquake and Tsunami, 24 May – 2 June
2011.
3. TEPCO Report to NISA (in Japanese).
4. Effects of the Earthquake and Tsunami on the Fukushima Daiichi and Daini Nuclear
Power Stations, May 24, 2011.
5. TEPCO Plant Data.
6. Various TEPCO, NISA, JNES and official press releases.
7. Japan Nuclear Technology Institute (JANTI). 24 March 2011. What happened in the
TEPCO Fukushima Daiichi Nuclear Power Station?
8. TEPCO Official Timeline.
9. Results of Monitoring at Fukushima Daiichi Nuclear Power Station.
Other Resources Researched During This Investigation:
Japan Atomic Industrial Forum, Inc. April 2011. Information on Status of Nuclear
Power Plants in Fukushima.
Japan Nuclear Energy Safety Organization (JNES). 19 April 2011. Assessment of theSpent Fuel Pool of Fukushima Daiichi Unit 4.
Nuclear and Industrial Safety Agency (NISA) and Japan Nuclear Energy Safety
Organization (JNES). 4 April 2011. The 2011 off the Pacific coast of Tohoku Pacific
Earthquake and the seismic damage to the NPPs.
Tokyo Electric Power Company. 9 April 2011. Investigation Results of Tsunami at
Fukushima Daiichi and Daini Nuclear Power Stations.
Tokyo Electric Power Company. 17 April 2011. Roadmap towards Restoration from the
Accident at Fukushima Daiichi Nuclear Power Station.
Tokyo Electric Power Company. 18 April 2011. The Great East Japan Earthquake andCurrent Status of Nuclear Power Stations.

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7.0

Additional Information
7.1

Earthquake Design Basis

The seismic design criteria for Japanese nuclear power plants is established by the
Nuclear Safety Commission (NSC) of Japan, as documented in NSC Regulatory Guides
for Power-generating Light Water Reactors. The seismic design criteria at Fukushima
Daiichi include geological survey information and calculate the design values based on
ground motion from known and hypothetical sources. The seismic design basis for
Fukushima Daiichi was derived from the 1938 Shioyazaki offshore earthquake, which
occurred in the vicinity of the site and was considered as a specified interplate earthquakethat had the most significant impact on the site. The hypothetical source is based on therelative uncertainty associated with geological conditions and the inability to fully
evaluate all probable earthquakes that could occur near the site. The probability ofexceeding the design basis seismic ground motion was calculated to be 10-4 to 10-6.
The March 11 earthquake occurred over the area where multiple smaller individual
earthquakes had previously occurred. The interaction over a large area contributed to theearthquake being the largest Japan has ever experienced and the fourth largest recorded
earthquake in the world. The design basis seismic analysis had not considered thepossibility of ground motion across several areas. The March 11 earthquake exceeded
the maximum acceleration value for units 2, 3, and 5 in the east-west direction, as
measured from the reactor building base slab seismometers.
The earthquake damaged breakers in the units 1 and 2 switchyard, causing a loss of offsite power to both units. A protective relay actuated, causing breakers in the Shin
Fukushima Power Substation to open, resulting in a loss of off-site power to units 3 and 4(the Unit 3 startup transformer was out of service for planned modification work before
the earthquake). Units 5 and 6 lost power when a transmission line tower that carriedboth 66-kV lines (tower Number 27) collapsed. As a result, the earthquake caused a lossof all off-site power to units 1 through 6.
While no formal seismic walkdowns have been performed, a review of plant indications
and operator logs does not indicate any seismic damage that affected the operator
response to the earthquake. Before the tsunami, all emergency diesel generators thatwere operable started and loaded as expected, and each emergency core cooling system
the operators used appeared to function as designed. Reactor pressure, reactor waterlevel, and containment pressure indications for units 1, 2, and 3 appeared as expectedfollowing a scram and did not indicate any potential breach of the reactor coolant systemfrom the earthquake.

45

INPO 11-005
Preliminary Seismic Data – Fukushima Daiichi reactor building slab seismic instrument3

Figure 7.1-1 Preliminary Seismic Data
Graphs indicate the observed values as compared to the design basis seismic criteria.
Ss-1: Design basis for inland crustal earthquakes and interplate earthquakes
Ss-2: Design basis for oceanic intraplate earthquakes
Ss-3: Design basis for hypothetical (unspecified) sources
3“Seismic Ground Motion Due to Great East Japan Earthquake and Seismic Ground Motion Accounted for in Seismic Safety Assessments,” provided by TEPCO

46

INPO 11-005
7.2

Tsunami Design BasisIn the 1960s, when TEPCO applied for the construction permit at Fukushima Daiichi, it
was common practice to adopt historical tsunami records as the design basis tsunami
height. Numerical simulation of tsunamis based on tsunami generation methods(earthquakes) did not begin until the mid-1970s. The original design basis tsunami for
Fukushima Daiichi was based on the Chilean tsunami of 1960, which resulted in ahistoric high water level of 10.2 ft (3.122 m) at the Onahama port, just north of the plant.
Based on this, 10.2 ft (3.122 m) was the design basis for Fukushima Daiichi when theconstruction permit was issued.
The tsunami design basis for Fukushima Daiichi considered only the inundation andstatic water pressures, and not the impact force of the wave or the impact of debris
associated with the wave. The design included a breakwater, which ranged in heightfrom 18 ft (5.5 m) to as high as 32.8 ft (10 m), as shown in Section 1.1.
Following the publishing of Tsunami Assessment Methods for Nuclear Power Plants inJapan by the Japan Society of Civil Engineers (JSCE) in 2002, TEPCO voluntarily
reassessed its tsunami design basis. Using these new deterministic evaluation techniques,
however, TEPCO determined the design basis tsunami would result in a maximum waterlevel of 18.7 ft (5.7 m). Because these changes were done voluntarily and not at thedirection of the regulator, the licensing basis did not change. According to theevaluation, the elevation of the Unit 6 seawater pump motor for the emergency diesel
generator was raised 7.9 in (20 cm), and the seawater pump motor for high pressure corespray was raised 8.7 in (22 cm). These changes ensured all vital seawater motors were
installed higher than the new inundation level of 18.7 ft (5.7m). The new analysis did notconsider or require the station design to mitigate hydrodynamic impact forces. The
breakwater was not modified when the new tsunami height was implemented because it
was not intended to provide tsunami protection, but rather to minimize wave action in the
harbor.
The 2002 analysis considered tsunamis generated from eight different near-field sources
off the coast of Japan. The March 11 earthquake was a rupture across several of theseareas, resulting in a larger-than-expected tsunami. Tsunamis caused by ruptures acrossseveral areas had not been considered as credible in the analysis. The tsunami was thelargest in the history of Japan.
In 2006, TEPCO performed a study on the development of probabilistic tsunami hazard
analysis, which used the Fukushima coast as an example. The study estimated theprobability of the Fukushima coast experiencing a tsunami greater than 19.7 ft (6 m) to beless than 1.0 E-2 in the next 50 years.

47

INPO 11-005

7.3

Main
Specifications

Station Specifications
Electric Output (MW)Unit 1
460

Unit 2
784

Unit 3784

Unit 4
784

Unit 5784

Unit 6
1,100

Start of Construction

Sep-67

May-69Oct-70

Sep-72

Dec-71

May-73

Start of Commercial Operation

Mar-71

Jul-74Mar-76

Oct-78

Apr-78Oct-79Reactor Type

BWR3

BWR4

BWR4
Mark I

BWR4

BWR4

BWR5
Mark II

Main Contractor

GE

GE/Toshiba

Toshiba

HitachiToshiba

GE/ToshibaHeat Output (MW)

1,380

2,381

3,293

Number of Fuel Assemblies

400

548

764

Full Length of Fuel Assemblies (in)

171

176

176

Number of Control Rods

97

137

185

Inner Diameter (in)

189

220

252

Height (in)

787

866

906

Total Weight (short ton)

485

Containment Type

Nuclear Reactor

ReactorPressure Vessel(RPV)

Primary
Containment
Vessel (PCV)

551
1194.81249.9Design Temperature (F)

572

576

Height (ft.)

105.0

108.3

Diameter of Cylindrical Portion (ft.)

32.8

36.1

32.8 (top)Diameter of Spherical Portion (ft.)

59.1

65.6

82.0 (bottom)

Suppression Pool Water Amount (kgal)

462.3

787.2

845.4

Design Pressure (psig)

62.4

Fuel
AC Distribution

111.5

157.5

55.1

40.6

284

Design Temperature (F)Number of Revolutions (rpm)Steam Turbine

827

Design Pressure (psi)280

340 (DW) 221 (SC)1,500

Steam Temperature (F)540

Steam Pressure (psig)

950

Type
Uranium (ton)69

Emergency Diesel Generators (* indicates air-cooled EDG)

2

Electrical Griduranium dioxide (Unit 3 contains MOX)
94
1/1*
4 - 275-kV lines

2

1/1*

2

132
2/1*
2 - 500-kV lines

Table 7.3-1: Fukushima Daiichi Unit Information (U.S. units)converted from source document provided in Table 7.3-2
48

INPO 11-005

Main
Specifications

Unit 1

Unit 2

Unit 3

Unit 4

Unit 5

Unit 6

Electric Output (MW)

460

784

784

784

784

1,100

Start of Construction

Sep-67

May-69Oct-70

Sep-72

Dec-71

May-73

Start of Commercial Operation

Mar-71

Jul-74Mar-76

Oct-78

Apr-78Oct-79Reactor Type

BWR3

BWR4

BWR4
Mark I

BWR4

BWR4

BWR5
Mark II

Main Contractor

GE

GE/Toshiba

Toshiba

HitachiToshiba

GE/ToshibaHeat Output (MW)

1,380

2,381

3,293

Number of Fuel Assemblies

400

548

764

Full Length of Fuel Assemblies (m)

4.35

4.47

4.47

Number of Control Rods

97

137

185

Inner Diameter (m)

4.8

5.6

6.4

Height (m)

20

22

23

Total Weight (metric ton)

440

Containment Type

Nuclear Reactor

ReactorPressure Vessel(RPV)

Primary
Containment
Vessel (PCV)

500
8.24

8.62

Design Temperature (℃)

300

302

Height (m)3233

Diameter of Cylindrical Portion (m)1011

10 (top)

Diameter of Spherical Portion (m)1820

25 (bottom)

Suppression Pool Water Amount (ton)

1,750

2,980

3,200

Design Pressure (MPa gauge)

0.43

Fuel
AC Distribution

34

0.28

140

138

Number of Revolutions (rpm)

1,500

Steam Temperature (℃)

282

Steam Pressure (kg/cm2g)

66.8

Type
Uranium (ton)69

Emergency Diesel Generators (* indicates air-cooled EDG)

2

Electrical Griduranium dioxide (Unit 3 contains MOX)
94
1/1*

2

1/1*

4 - 275-kV lines

Table 7.3-2: Fukushima Daiichi Unit Information (metric units)4

4

Overview of facility of Fukushima Daiichi Nuclear Power Station, http://www.tepco.co.jp/en/nu/fukushima-np/index-e.html

49

48

0.38

Design Temperature (℃)
Steam Turbine

750

Design Pressure (MPa)2171 (DW) 105 (SC)

132
2/1*
2 - 500-kV lines

INPO 11-005
Unit 1

Unit 2

Unit 3

Number of systems

2

2

2

Flow (gpm per system)

2422

4491

5024

Number of pumps per system

2

1

1

Pump discharge pressure (psig)

284

501

501

Number of systems

2

2

2

Flow (gpm per system)

3104

13032

11447

Number of pumps per system

2

2

2

Number of heat exchangers per system

1

1

1

High Pressure
Coolant
Injection(HPCI)

Number of systems

1

1

1

Flow (gpm per system)

3003

4249

4249

1

1

1

Low Pressure
Coolant
Injection (LPCI)

Number of systems

2

2

Flow (gpm per system)

7705

8013

2

2

4

4

Flow (gpm)

7705

8013

Total pump head (ft)

420

420

4

4

Flow (gpm)

4306

4306

Total pump head (ft)

761

761

2

2

7760000

7760000

Core Spray(CS)

Containment
Cooling (CCS)

Number of pumps per system

Number of pumps per system
Pump
Number of pumps

Residual Heat
Removal (RHR)

Seawater pump
Number of seawater pumps

Heat exchanger
Number of units
Heat transfer capacity (kcal/hr)
Pump
2

Number of pumps
Shutdown
Cooling (SHC)

Flow (gpm)

2050

Pump head (ft)150

Heat exchanger
Number of heat exchangers

2

Heat exchanging capacity (kW)4400

Table 7.3-3: Fukushima Daiichi System Information (U.S. units)converted from source
document provided in Table 7.3-4

50

INPO 11-005
Unit 1

Unit 2

Unit 3

Number of steam turbines

1

1

Reactor pressure (psig)

1045-150

1045-150

Output (HP)

500-80500-80Speed (rpm)5000-2000

5000-2000

Number of pumps

1

1

Flow (gpm)

418

427

Total pump head (ft)2788-525

2788-525

Speed (rpm)

Variable

Variable

Steam turbine

Reactor Core
Isolation
Cooling (RCIC)

Isolation
Condenser (IC)Standby Gas
Treatment(SGTS)

Safety Valves

Pump

2

Number of systems
Tank water retention capacity (gal/tank)

28002

Steam flow (short ton/hr per tank)

116

Number of systems

2

2

2

Number of fans per system

1

1

1

Exhaust capacity (cfm per unit)

1101

1589

1589

Iodine filtration efficiency (%)

>97

>99.9

>99.9

Number of valves

3

3

3

Total capacity (short ton/hr)

992

992

992

1241.11241.11235.41251.1

Blowout pressure (psig)
Blowoff area

(2 valves)(1 valve)

Drywell

Drywell

4

8

8

1202

3197

3197

Number of valves
Total capacity (short ton/hr)

Main Steam
Safety Relief
ValvesRelief valve function (psig)

Safety valve function (psig)
Blowoff area

DrywellPressure

ValvesPressure

ValvesPressure

Valves1056.111080.311080.311066.121090.2

31090.2

3

1076.01

1100.2

4

1100.241110.221110.221120.1

21120.131130.1

3

Suppression Chamber

Table 7.3-3 (continued)

51

Suppression Chamber

Suppression Chamber

INPO 11-005
Unit 1

Unit 2

Unit 3

Number of systems

2

2

2

Flow (T/hr per system)

550

1020

1141

Number of pumps per system

2

1

1

Pump discharge pressure (kg/cm2g)

20

35.2

35.2

Number of systems

2

2

2

Flow (T/hr per system)

705

2960

2600

Number of pumps per system

2

2

2

Number of heat exchangers per system

1

1

1

High Pressure
Coolant Injection(HPCI)

Number of systems

1

1

1

Flow (T/hr per system)

682

965

965

Number of pumps per system

1

1

1

Low Pressure
Coolant Injection(LPCI)

Number of systems

2

2

Flow (T/hr per system)

1750

1820

2

2

4

4

Flow (T/hr)

1750

1820

Total pump head (m)

128

128

Number of seawater pumps

4

4

Flow (m3/hr)

978

978

Total pump head (m)

232

232

2

2

7760000

7760000

Core Spray (CS)Containment
Cooling (CCS)

Number of pumps per system
Pump
Number of pumps

Residual Heat
Removal (RHR)

Seawater pump

Heat exchanger
Number of units
Heat transfer capacity (kcal/hr)
Pump
2

Number of pumps
Shutdown
Cooling (SHC)

Flow (m3/hr)

465.5

Pump head (m)45.7Heat exchanger
2

Number of heat exchangers

3800000

Heat exchanging capacity (kcal/h)Table 7.3-4: Fukushima Daiichi System Information (metric units)55

Report of the Japanese Government to the IAEA Ministerial Conference on Nuclear Safety, June 201152

INPO 11-005
Unit 1

Unit 2

Unit 3

Number of steam turbines

1

1

Reactor pressure (kg/cm2g)

79-10.6

79-10.6

Output (HP)

500-80500-80Speed (rpm)5000-2000

5000-2000

Number of pumps

1

1

Flow (t/h)

95

97

Total pump head (m)850-160

850-160

Speed (rpm)

Variable

Variable

Steam turbine

Reactor Core
Isolation Cooling(RCIC)

Isolation
Condenser (IC)Standby Gas
Treatment(SGTS)

Safety Valves

Pump

Number of systems

2

Tank water retention capacity (m3/tank)106

Steam flow (metric ton/hr per tank)

100.6

Number of systems

2

2

2

Number of fans per system

1

1

1

Exhaust capacity (m3/hr per unit)

1870

2700

2700

Iodine filtration efficiency (%)

>97

>99.9

>99.9

Number of valves

3

3

3

Total capacity (metric ton/hr)

900

900

900

87.2

87.2

Drywell

Drywell

Drywell

4

8

8

1090

2900

2900

(2
valves)(1
valve)

86.8
87.9

Blowout pressure (kg/cm2g)

Blowoff area
Number of valves
Total capacity (metric ton/hr)

Main Steam
Safety Relief
ValvesPressure

ValvesPressure

ValvesPressure

Valves

74.2

1

75.9

1

75.9

1

74.9

2

76.6

3

76.6

3

75.6

1

77.3

4

77.3

4

78

2

78

2

78.7

2

78.7

3

79.4

3

Relief valve function (kg/cm2g)

Safety valve function (kg/cm2g)
Blowoff area

Suppression Chamber

Table 7.3-4 (continued)

53

Suppression Chamber

Suppression Chamber

INPO 11-005
7.4

Fukushima Daiichi Simplified System Drawings/Descriptions

MO
2

MO
2

B

A

Isolation
Condenser B

Primary Containment Vessel

Isolation
Condenser A

MO
1

MO
1

B

RPV

A

RR Pump (B)

MO
3

MO
4

MO
3
B

B

MO
4
A

Figure 7.4-1 Isolation Condensers on Unit 16

6

Report of the Japanese Government to the IAEA Ministerial Conference on Nuclear Safety, June 2011

54

A

INPO 11-005

Primary Containment Vessel

Unit 1 is 4

RPV

MS(C)

Unit 1 is 5

MO
15

MO
16

FDW(B)

Unit 1 is Z8
Governing Valve

Unit 1 is 7

AO
18

Unit 1 is 8MO
19

Unit 1 is Z9

Unit 1 is 3

Stop Valve

HO

HO

*

*

MO
14

EGR

S/C
Unit 1 is 35

Unit 1 is 9

MO
20

HPCI
Booster Pump

HPCI
Control System

MO
57

MO
17

HPCI
Main Pump

HPCI Turbine

Exhaust (To Suppression Chamber)
CST
Unit 1 is 6

Figure 7.4-2 High Pressure Coolant Injection (HPCI) – Typical Arrangement, Units 137

7

Report of the Japanese Government to the IAEA Ministerial Conference on Nuclear Safety, June 2011

55

INPO 11-005

Primary Containment Vessel

MO
15

MO
16

RPV

MS(B)

Unit 3 is 10-8

Unit 3 is Z1

FDW(A)

HO
Z14

HO
Z14

MO
131

AO
22
EGR
MO
21

S/C

MO
39

MO
20

RCIC
Control System

RCIC Pump
RCIC Turbine

MO
18

Exhaust (To Suppression Chamber)
CST

Figure 7.4-3 Reactor Core Isolation Cooling (RCIC) – Typical Arrangement, Units 2 and 38

8

Report of the Japanese Government to the IAEA Ministerial Conference on Nuclear Safety, June 2011

56

INPO 11-005

Primary Containment Vessel Vent System Description
Between 1999 and 2001, TEPCO modified the design of the containment vent system byadding new vent pipes extending from the suppression chamber and drywell to the stacks.
These vent pipes were added to provide a method of venting containment during severe
accidents. The modified ventilation piping bypassed the standby gas treatment system(SGTS) to allow operators a method of venting containment even when the pressure ishigh. The piping connection with the ventilation stack is provided with a rupture disk.
(See figures 7.4-4 and 7.4-5.)

Figure 7.4-4 Overview of Unit 1 Containment Vent System99Report of the Japanese Government to the IAEA Ministerial Conference on Nuclear Safety, June 2011

57

INPO 11-005

Figure 7.4-5 Overview of Units 2 and 3 Containment Vent System10

10Report of the Japanese Government to the IAEA Ministerial Conference on Nuclear Safety, June 201158

INPO 11-005
Fukushima Daiichi Units 1 through 4 before the event

Figure 7.4-6 Simplified Electrical System11
Systems highlighted in green were in service (energized). The 4A EDG and Unit 3 startup transformer were out of service for planned
maintenance at the time of the event.
11“Fukushima Nuclear Accident Interim Report,” September 13, 2011, Tokyo Electric Power Company

59

INPO 11-005
Fukushima Daiichi
Unit 1

Unit 6

StatusEquipment
Name

StatusEquipment
Name

StatusEquipment
Name

StatusEquipment
Name

StatusEDG 1Ax

EDG 2Ax

EDG 3Ax

EDG 4Ax

EDG 5A(2)

EDG 6A(2)

EDG 1Bx

EDG 2B(ac)(1)

EDG 3Bx

EDG 4B(ac)(1)

EDG 5B(2)

EDG 6B
(ac)

o

--

--

--

--

--

--

--

--

--

--

HPCS EDG

(2)

Vital

M/C 1Cx

M/C 2Cx

M/C 3Cx

M/C 4Cx

M/C 5Cx

M/C 6Co

M/C 1Dx

M/C 2Dx

M/C 3Dx

M/C 4Dx

M/C 5Dx

M/C 6Do----

M/C 2Ex----

M/C 4Ex----

HPCS M/Co

M/C 6A-1

x

M/C 6A-2

x

M/C 6B-1

x

M/C 6B-2

x

Non-Vital

M/C 1Bxx

M/C 2A
M/C 2B
M/C 2SA

Vital

x
x
x

M/C 3A
M/C 3BM/C 3SAxx

M/C 4Ax

M/C 4Bxxx-M/C 2SBx

M/C 3SBx

M/C 5Ax

M/C 5Bx

M/C 5SA-1

x

M/C 5SA-2

x

M/C 5SB-1

x

M/C 5SB-2

x--

P/C 1Cx

P/C 2CP/C 3Cx

P/C 4CoP/C 5Cx

P/C 6CoP/C 1Dx

P/C 2DP/C 3Dx

P/C 4Do

P/C 5Dx

P/C 6Do----

P/C 2E----P/C 4Ex----

P/C 6Eo

P/C 3Ax

P/C 4Ao

P/C 5Ax

P/C 6A-1

x

--

--

--

--

P/C 5A-1

o

P/C 6A-2

x

P/C 3Bx

P/C 4BoP/C 5Bx

P/C 6B-1

x

x

P/C 2Ax

P/C 1Bx

P/C 2B

--

--

--

--

--

--

--

--

P/C 5B-1o

P/C 6B-2

x

P/C 1Sx----

P/C 3SAx----

P/C 5SA

x

--

--

--

--

--

--

--

--

P/C 5SA-1

x

--

--

-DC 125V
main busA
DC 125V
main bus
B--

P/C 2SBx

P/C 5SBx----

DC 125V
P/C 2Ax

x

DC 125VP/C 5Ao

DC 125V
6Aox

DC 125V
P/C 2Bxx

DC 125VP/C 5Bo

DC 125V
6B

o

SW

x

RHR-S Ax

x

-DC 125V
main bus
4A
DC 125V
main bus
4B
RHR-S A--x

P/C 3SB
DC 125V
main bus
3A
DC 125V
main bus
3B
RHR-S Ax

RHR-S Ax

RHR-S Ax

RHR-S Bx

RHR-S Bx

RHR-S Bx

RHR-S Bx

RHR-S Bx

Non-Vital

P/C 1A

UHS

125V

Unit 5Equipment
Name

M/C 1S

480V Power Centers (P/C)

Unit 4

StatusM/C 1ADC Power

Unit 3Equipment
NameEDG(ac) = air cooled
6.9 kV Electrical Distribution

Unit 2

Status:Key:

P/C 2A-1

x

x
o

o

x: damaged
o: available
White background: Not damaged by the earthquake or tsunami
Blue background: Damaged or flooded by tsunami
Gray background: Support systems damaged or flooded by tsunami(1): electrical distribution damaged or flooded(2): ultimate hat sink damaged or flooded
Figure 7.4-7 Fukushima Daiichi Electrical Distribution Damage1212“Overview of Accident at TEPCO Fukushima Nuclear Power Stations,” July 22, 2011 - Tokyo Electric Power
Company Co.

60

INPO 11-005

Figure 7.4-8: Generic cross-section of a BWR4 with a Mark I containment (similar to
Fukushima Daiichi)

61

INPO 11-005
7.5

Organizational Structure and Staffing

Fukushima Daiichi uses one shared control room for each two units on site (three control
rooms; five shift crews per control room). One operating crew is responsible for eachcontrol room and the two associated units. The crew rotation is a standard five-crew, 12hour shift rotation with 10 days of training after every fifth rotation. The chain ofcommand is configured in the following manner (units 1-2 shown).

Site Superintendent

Unit Superintendent (Units 1-4)

Unit Superintendent (Units 5-6)

Operations Department General
Manager (Units 1-4)

Operations Department General
Manager (Units 5-6)

Shift Supervisor

Assistant Shift Supervisor

Unit 1 Senior
Operator

Unit 2 Senior
Operator

Assistant Senior
Operator
Unit 1 Main Shift Operator

Auxiliary Operator

Auxiliary Operator

Unit 2 Main Shift Operator

Auxiliary Operator

62

Auxiliary Operator

INPO 11-005
Approximate equivalent positions for U.S. nuclear units (not exact equivalent):Japan
shift supervisor
assistant shift supervisor
unit senior operator
unit main shift operator
assistant senior operator
auxiliary operator

U.S.
shift manager
control room supervisor
at-the-controls operator (similar to senior reactor
operator level in Japan)
balance-of-plant reactor operator (reports to unit
senior operator)
field supervisor
nonlicensed operator

During severe accident conditions, the site supervisor is in charge of site Emergency
Response Center supervision. The shift supervisors will report directly to the site
superintendent during these conditions.
Operations Staffing at the Time of the Event
The operating crews consisted of the following:

Units 1 and 2 had 11 operators and 1 trainee.

Units 3 and 4 had 8 operators and 1 trainee. (Unit 4 minimum shift staffing was
reduced because of the refueling outage.)The normal operating shift organization for each of the two unit crews includes
one shift supervisor, one assistant shift supervisor, two senior operators, one
assistant senior operator, two main shift operators, and four auxiliary operators.



One crew in training reported to the station to assist in the response.

Station Staffing at the Time of the Event

7.6



Immediately after the tsunami, approximately 400 people (about 130 operatorsand 270 maintenance personnel) were available for the recovery process.



About 70 TEPCO employees (maintenance workers) and about 40 people fromaffiliated companies were engaged in the initial field work to recover units 1
through 3. Most of the work was recovery of instrumentation and power supplies.
Operator Training

Initial operator training programs for reactor operators and above are provided by BWR
Training Center Corporation (BTC), which is located close to Fukushima Daiichi.
TEPCO is one of the shareholders of the company. The site superintendent certifies
initial qualifications for senior operators, main shift operators, and auxiliary operators onsite. By law, shift supervisors are required to be licensed. The licensing process includes

63

INPO 11-005
an evaluation of their knowledge and skills. The Japan Nuclear Technology Institute(JANTI) performs this evaluation and issues the license.
Continuing training programs are developed and implemented at the station. Operators
receive approximately 80 hours of continuing training per year. A training curriculum
review committee selects the training subject material for continuing training. The
training topics are typically based on operator fundamentals, performance improvement,
operating experience, and changes in job performance requirements. Training is
performed in the classroom and simulator, as well as in on-the-job training settings.
Some training is performed for specific positions, such as the shift supervisor (SS), senior
operator (SO), and main shift operator (MSO). Continuing training topics include thefollowing:
Topic
Population
Team Operation (Reactor Startup, AOP/EOP)ALL
Plant Systems, AOP/EOP
MSO/SO
AOP/EOP/SOPSS
Fundamentals
ALL
Technical Specifications
ALL
Human Performance, Expectations
ALL

Location Time (hours)
SIM
30
Class/SIM
24
Class/SIM
3
Class
9
Class
9
Class
3

During continuing training, all operators are required to maintain and enhance theirperformance to support safe and reliable operation. To ensure this goal is met, operator
knowledge and skills are evaluated through written examinations, simulator evaluations,and task-performance evaluations. The examination standard includes requirements forhigher-order learning objectives, such as demonstrating the ability to diagnose plantconditions and prioritize response actions.
Fukushima Daiichi has one full-scope simulator, which models Unit 3 (BWR-4 with aMark I containment); and a limited-scope simulator, which models Unit 1 (BWR-3 with aMark I containment). Operators in TEPCO also use two full-scope simulators at the
BTC, which models units 3 and 4 (BWR-4 with a Mark I containment).
The diagram shown below illustrates the typical progression and training requirementsfor operators. The times shown indicate the approximate time spent in each position.

64

INPO 11-005

Figure 7.6-1 Operator Progression and Training Requirements
7.7

Regulatory Structure

Governmental responsibility for the safe operation of Japan's nuclear power plants isdivided between multiple government agencies. These agencies, and their
responsibilities, are as follows:

The Minister of Economy, Trade and Industry (METI) has jurisdiction overcommercial nuclear power reactor facilities in Japan. The Nuclear and Industrial
Safety Agency (NISA), which operates under the authority of METI, is
responsible for the safety of nuclear energy. NISA’s mission is to ensure thesafety of the people’s livelihoods through the regulation of the energy industry
and related industries.



In October 2003, the Japan Nuclear Energy Safety Organization (JNES) wasestablished. JNES inspects nuclear facilities and provides technical support safety
reviews and assessments.



The Ministry of Education, Culture, Sports, Science and Technology (MEXT) has
various responsibilities, including environmental radiation monitoring. MEXT is
divided into two basic groups: the research and development bureau and thescience and technology policy bureau. The research and development bureau is
responsible for the promotion of nuclear energy, and the science and technology

65

INPO 11-005
policy bureau is responsible for nuclear safety regulation. MEXT is also
responsible for dose limits for occupational exposure and helps local governments
perform environmental surveys following an accident.


The Nuclear Safety Commission (NSC) is an independent agency that operates
under the cabinet office. The NSC is an administrative organization that provides
supervision and audits the safety regulation by the MEXT and METI, thus
providing independent verification over the administration of nuclear regulationsin Japan. The NSC has a range of missions, including planning, deliberation, andmaking decisions on regulations and policies related to nuclear safety as well as
prevention of radiation hazards, based on expert knowledge on nuclear
technologies and radiological protection. The NSC has the authority to makerecommendations to the regulatory bodies via the Prime Minister. The NSC also
promotes communications with the public for enhancing the credibility of nuclear
safety.

NISA occupies an off-site center, which must be located within 10 km of the station.
Each nuclear station is required to have its own center, even when stations are in close
proximity. These centers are equipped to monitor plant and meteorological conditions.
The JNES analyzes the data and makes recommendations on sheltering or evacuation.
JNES makes the recommendation to NISA, which takes that recommendation and passes
it along to the various government organizations.
Emergency Plan
Emergency preparedness and implementation in Japan involve many organizations,
including on site, off site, national government, local government(s), regulators, andcontractors. Processes, organizations, and activities include responses to reportable
events and nonreportable emergencies, as defined below. A drawing is provided todemonstrate some of these relationships.
One large-scale national emergency response drill is conducted each year to exercise theassociated organizations and ensure the emergency plan is capable of fulfilling its
intended function.

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INPO 11-005

Figure 7.7-1: Nuclear Emergency Response Organization Relationships13
Nuclear Emergency Preparedness Act
The Act on Special Measures Concerning Nuclear Emergency Preparedness (commonlyreferred to as the Nuclear Disaster Law) was established in 1999 in response to the
September 30, 1999 inadvertent criticality accident at the Tokai uranium processing
plant. The accident resulted in overexposure of three plant workers and additional
unplanned exposures to 66 plant workers, local inhabitants, and emergency support
personnel.
The Nuclear Emergency Preparedness Act provides guidance for responding to a nuclear
emergency. The act is intended to provide a closely coordinated response among therelevant organizations. It includes the following guidance:
13

Report of the Japanese Government to the IAEA Ministerial Conference on Nuclear Safety, June 2011

67

INPO 11-005
1) Licensee notification to the Minister of METI is required following declaration ofa “special event” as stipulated in Article 10 of the Nuclear Emergency
Preparedness Act.
2) Upon receiving notification, the Minister shall initiate required actions andsupport local governments to prevent further escalation of a disaster.
3) If the Minister recognizes that a “special event” has degraded and conditions
exceed predetermined thresholds associated with a nuclear emergency situation,
the Minister shall immediately notify the Prime Minister.
4) The Prime Minister is responsible for declaring a “nuclear emergency situation”
and directing local governments to take protective measures, such as evacuation,
sheltering, and the administering of potassium iodide.
5) The Prime Minister shall establish and head the Nuclear Emergency Response
Headquarters (NERHQ) in Tokyo and establish the local NERHQ at theassociated off-site location.
6) The NSC shall convene an organization composed of commissioners andadvisors, known as the Technical Advisory Organization in an Emergency, which
will provide technical advice to the Prime Minister.
7) Local governments shall establish their own emergency response headquarters.
8) The Joint Council for Nuclear Emergency Response shall be established to share
information between national and local governments and related organizations.
TEPCO also has a corporate emergency response manual. This manual defines threealert conditions for when a disaster occurs or is likely to occur. The station emergency
response organization is temporarily activated when these alert conditions are declared.
The three levels of alert are as follows:Alert level 1 is declared when a disaster is predicted or has occurred.
Alert level 2 is declared when a large-scale disaster is predicted or has occurred.
Alert level 3 is declared when a large-scale disaster has occurred that may requirea lengthy recovery.

The emergency response organization works out of a station Emergency Response
Center. This is a new building that was added following the Niigata-Chuetsu-oki
earthquake in 2007. The building was seismically designed, is at an elevated location,and has backup generators to provide power, so it withstood both the earthquake and thetsunami. This is the work location for the site superintendent when fulfilling the role ofemergency director.

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INPO 11-005
Following an emergency declaration, the corporate office will activate the corporate
emergency response center in Tokyo. The corporate emergency response center isstaffed to advise and instruct the station as needed and to support with media relations.
The media release briefing room is located just outside the TEPCO Head Office Nuclear
Emergency Headquarters.
Station Emergency Response Organization Training
According to TEPCO, the station emergency response organization conducts several
integrated exercises each year. The number of scenarios is sufficient to ensure that every
member of the emergency response organization participates in at least one drill per year.
Lower-level or specific drills, such as emergency medical treatment, evacuations, andcommunications exercises, are also performed throughout the year. The operators will
also participate in at least one emergency response organization emergency preparedness
drill each year in the simulator.

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7.8

Acronyms and Abbreviations
Abbreviations

Units of Measure

AMG

accident management guidelines

abs

absolute

AOP

abnormal operating procedure

AC

Alternating Current

BWR

boiling water reactor

Bq

Becquerel

EDG

emergency diesel generator

C

CelsiusEOP

emergency operating procedurecm

centimeters

ERC

Emergency Response Center

cpm

counts per minute

HPCI

high pressure coolant injection

DC

Direct Current

IAEA

International Atomic Energy Association

F

FahrenheitICisolation condenser

ft

feet

INES

International Nuclear and Radiological Event Scalegal

galileo (seismic intensity)

INPO

Institute of Nuclear Power Operations

gal

gallon (volume)

JANTI

Japan Nuclear Technology Institutegpm

gallons per minute

JNES

Japan Nuclear Energy Safety Organization

km

kilometers

JST

Japan Standard TimekPa

kilopascals

M/C

metal clad switchgearkVkilovolts

METI

Ministry of Economy, Trade and Industry

lpm

liters per minute

MEXT

Ministry of Education, Culture, Sports, Science and Technology

MPa

Megapascals

MSO

main shift operator

MWemegawatts electric

NERHQ

Nuclear Emergency Response HeadquartersmmetersNISA

Nuclear and Industrial Safety Agency

uSv

microsievert

NPP

nuclear power plantmm

millimeters

NSC

Nuclear Safety Commission

mremmillirem

P/C

power center

mSv

millisievert

PCV

primary containment vesselpsipounds per square inch

RCIC

reactor core isolation coolingpsia

pounds per square inch absolute

RHR

residual heat removal

psig

pounds per square inch gauge

RPV

reactor pressure vessel

SFP

spent fuel pool

SGTS

standby gas treatment systemSIM

simulator

SLC

standby liquid control

SOsenior operator

SOP

standard operating procedure

SRO

senior reactor operator

SRV

safety relief valveSSshift supervisor

TAF

top of active fuelTEPCO

Tokyo Electric Power Company

UHS

ultimate heat sink70

Unit 1 Validated Event Timeline
8.0

INPO 11-005

Event Progression and Timeline
8.1

Unit 1Date

Time

11-Mar
11-Mar
11-Mar
11-Mar
11-Mar
11-Mar
11-Mar
11-Mar
11-Mar
11-Mar1446

11-Mar1447

11-Mar1452

11-Mar1452

11-Mar
11-Mar1502

11-Mar1503

11-Mar1506

11-Mar

1507

1446
1446
1447
1447
1447
1447
1447
1447
1447

1503

Unit 1 Actions
Ground motion exceeds the setpoint of seismic sensor C.
Ground motion exceeds the setpoint of seismic sensor B.
Automatic reactor scram signal on seismic trip
Plant computer indicates all control rods have fully inserted.
Inboard and outboard main steam isolation valves close because of loss of off-site power.
Automatic turbine trip on high vibration
6.9-kV bus 1D power loss
6.9-kV bus 1C power loss
EDG 1B starts and loads, energizing the 6.9-kV bus 1D; 6.9-kV bus 1D power loss clears.
EDG 1A starts and loads, energizing the 6.9-kV bus 1C; 6.9-kV bus 1C power loss clears.
Reactor water level initially dropped because of the collapsing steam voids. Reactor water level was within thenormal band and the operators did not need to initiate high pressure coolant injection.
The reactor mode switch was in shutdown.
Both isolation condensers (ICs) started automatically. Reactor pressure vessel (RPV) pressure begins to decrease ascooler water is recirculated through the reactor core from the IC. Reactor water level was in the normal level range,and high pressure coolant injection (HPCI) was not required to control level.
The reactor was confirmed subcritical.
The operators manually removed the isolation condensers from service.
Workers removed the IC from service by closing the cold leg return drywell outboard isolation motor-operated valves(MO-3A and B). These actions were taken because cooldown rate was exceeding the 100oF/hr (55oC/hr) limit
specified in technical specifications.
The Headquarters for Major Disaster Countermeasures was established in the TEPCO office in Tokyo to assess
damages from the earthquake and recover from the electric outage in TEPCO's service area.
Torus cooling and spray were reported in service (1507-1510).

71

Unit 1 Validated Event Timeline
Date

Time

11-Mar1510

11-Mar
11-Mar
11-Mar
11-Mar
11-Mar1527

11-Mar--

11-Mar--

11-Mar

1542

11-Mar--

11-Mar

1636

1535
1537
---

11-Mar--

11-Mar--

11-Mar1645

INPO 11-005

Unit 1 Actions
Operators determined that only one train of IC was needed to control reactor pressure in the 870 and 1,015 psi (6 to
7 MPa). The A IC system was operated by the opening and closing of the motor-operated valve (MO-3A) to control
reactor pressure.
The first wave of a series of tsunamis, generated by the earthquake, arrived at the station.
The second tsunami hit the station.
Loss of all AC power occurs. Instrumentation and emergency systems gradually fail between 1537 and 1550.
The loss of DC distribution systems results in the loss of control room indications and alarms.
The control room lighting was lost and only emergency lighting remained.
The control panel indications for HPCI were barely lit but slowly faded to black. The operators determined HPCI was
not operable because indicators on the control panel had gone out.
Valve status of IC was lost on the control panel.
TEPCO entered its emergency plan because of the loss of all AC power, in accordance with Article 10, paragraph 1 of
the Nuclear Disaster Law. Government offices were notified. The corporate Emergency Response Center was
established.
Flooding caused a loss of the instrumentation power system, resulting in a major loss of control room indications,
including the reactor water level indication
Operators could not determine reactor water level or the status of injection into the reactor. An emergency was
declared because of the loss of emergency core cooling system injection sources, in accordance with Article 15,
paragraph 1 of the Nuclear Disaster Law. The appropriate government agencies were notified at 1645.
Temporary batteries and cables were gathered and carried to the units 1 and 2 control room. After confirming thewiring layout using drawings, batteries were connected to instrument panels. The top priority was to verify thestatus of water injection into the RPV, so efforts were focused on connecting batteries to the DC-powered reactor
water level indicator.
The Emergency Response Center (ERC) began reviewing the accident management procedures and checking the vent
procedures to determine how to open the containment vent valves without power. Workers in the ERC went
through the administration building and retrieved drawings and manuals needed to develop the procedure.
Operators were able to determine reactor water level. The emergency associated with the inability to determine
reactor water level was cancelled, and the appropriate government agencies were notified at 1655.

72

Unit 1 Validated Event Timeline
Date

Time

11-Mar1707

11-Mar1712

11-Mar--

11-Mar1730

11-Mar--

11-Mar--

11-Mar1818

11-Mar1818

11-Mar1825

11-Mar

2007

11-Mar2049

11-Mar2050

11-Mar2119

11-Mar2123

11-Mar2130

11-Mar
11-Mar

2151
2200

INPO 11-005

Unit 1 Actions
The operators again lost the ability to verify reactor water level or the status of water injection. The emergency plan
was reentered in accordance with Article 15, paragraph 1 of the Nuclear Disaster Law. The appropriate government
agencies were notified at 1712.
The site superintendent directed workers to investigate methods of injecting water into the reactor using the fire
protection system or fire trucks.
Operators reviewed methods for alternative water injection using the accident management (AMG) procedures andconfirmed the injection path using the installed fire pump into the RPV via the core spray system.
The installed diesel-driven fire pump was started and allowed to idle in standby.
Operators lined up the alternative water injection valves from the fire protection system to the core spray system bymanually opening the valves in the reactor building. The work was performed in complete darkness.
Injection could not be achieved until after the RPV was depressurized below 100 psi (0.69 MPa).
Somehow, DC power partially returned, which allowed the control board indication lights for the IC M0-3A and M02A to energize. The valves indicated closed.
Operators placed the isolation condenser in service by opening the MO-3A and MO-2A. Steam was observed comingfrom the condenser.
The isolation condenser was removed from service by the closing of the MO-3A valve.
Because there were no working indications in the control room, operators checked reactor pressure locally in thereactor building. Reactor pressure was 1,000 psi (6.9 MPa).
A small portable electric generator was installed, which restored some temporary lighting in the units 1-2 control
room.
Authorities of the Fukushima prefecture ordered evacuation of the population within a 1.2 mile (2 km) radius of
Fukushima Daiichi.
Water level indication was restored in the control room. Reactor water level was approximately 8 inches (200 mm)
above the top of active fuel (TAF).
The Prime Minister ordered the evacuation of the population within a 1.9 mile (3 km) radius of the station andsheltering for the population within a 1.9 to 6.2 mile (3 km to 10 km) radius of the station.
Operators placed the isolation condenser in service by opening valve MO-3A. The generation of steam wasconfirmed at the condenser.
Access to the reactor building was restricted because of high dose rates.
Reactor water level indicates 21.7 inches (550 mm) above the top of active fuel.

73Unit 1 Validated Event Timeline
Date

Time

11-Mar2300

11-Mar2350

12-Mar--

12-Mar0006

12-Mar--

12-Mar0030

12-Mar0049

12-Mar0130

12-Mar0145

12-Mar0148

12-Mar--INPO 11-005

Unit 1 Actions
A radiation survey identified dose rates of 120 mrem/hr (1,200 µSv/hr) in front of the reactor building north door on
the first floor of the turbine building and 50 mrem/hr (500 µSv/hr) in front of the door to the south. The government
authorities were notified at 2340.
A restoration team from the ERC connected the temporary generator, used to provide some control room lighting, tothe drywell pressure instrument. The indication read 87 psi (600kPa).
Late in the evening on March 11 and into the early morning of March 12, power supply vehicles from Tohoku ElectricPower Company and TEPCO arrived at the site.
The site superintendent directed preparations to vent the primary containment vessel (PCV) because drywellpressure was likely to exceed 87 psi (0.6 MPa).
In the control room, operators assembled piping and instrumentation drawings, the accident management
procedures, valve drawings, and a white board. The operators began to develop a procedure for venting, including
how to manually operate the valves, and the associated sequence.
The government confirmed the evacuation of the population within 1.9 miles (3 km) of the station.
An emergency declaration was made for an abnormal increase in drywell pressure because drywell pressure was
likely to exceed 87 psi (0.6 MPa). The appropriate government agencies were notified at 0055.
At approximately 0130, TEPCO officials informed the Prime Minister, Minister of METI, and NISA of their proposal tovent the containment. The Prime Minister, Minister of METI, and the NISA concurred. The TEPCO corporate ERC told
the station that at 0300, the Minister of METI and other associations would announce the venting; and that after theannouncement, the station should vent.
The government reconfirmed the evacuation of the population within 1.9 miles (3 km) of the station.
At some point, the installed diesel-driven fire pump that was standing by to pump water into the reactor
malfunctioned. In an attempt to restart the fire pump, diesel fuel was carried to the pump and the fuel tank wasrefilled, and batteries stored in an office were carried to the room and installed; but the pump would notstart. Workers began considering using fire trucks to supply water to the plant fire protection system.
Damage to fire hydrants and leaks from the fire system made it impossible to use filtered water as a water source forthe fire engines to supply the RPV. Isolation valves were closed to stop the leaks and maintain water level in thefiltered water tanks. The fire protection tank was confirmed available for use as a water source.

74Unit 1 Validated Event Timeline
Date

INPO 11-005

Unit 1 Actions

Time

12-Mar--

12-Mar--

12-Mar--

12-Mar0224

12-Mar
12-Mar

0230

The station had three fire engines, but only one was available to support injecting water into the Unit 1 RPV. One fireengine was damaged by the tsunami. The second fire engine was at parked adjacent to units 5 and 6 but could not
be driven to Unit 1 because earthquake damage to the road and debris from the tsunami had restricted accessbetween units 1 through 4 and units 5 and 6.
To use the available fire engine, workers had to clear obstacles and debris to move the fire engine close to Unit 1. A
heavy fuel oil tank, which had been displaced by the tsunami, blocked the road in front of the old administrationbuilding, making the road impassable. The security gate had lost power and would not open, preventing access fromthat road. The emergency response team broke a lock on the gate between units 2 and 3, allowing the fire engine toarrive at Unit 1.
Alternatives for injecting water via a fire protection line were reviewed. Additional fire engines and watertransportation by the Self-Defense Force were also considered.
In preparation for manually venting the containment, a radiological evaluation of working conditions in the torus
room was provided to the ERC. With radiation levels at 30 rem/hr (300 mSv/hr), workers were limited 17 minutes oftime in order to remain below the emergency response radiation limit of 10 rem (0.1 Sv). Workers were required towear a self-contained breathing apparatus (SCBA) with a 20-minute air supply and would be given potassium iodide
tablets.
Containment pressure had increased to 122 psia (0.84 MPa abs). The government was informed at 0247.

0245

Reactor pressure decreased to 116 psi (0.8 MPa).

12-Mar0255

12-Mar0306

12-Mar0345

12-Mar--

12-Mar
12-Mar

0401
0419

Unit 2 field workers returned to the control room and reported that RCIC was in operation on Unit 2. Based on thisreport, venting the Unit 1 containment was given a higher priority.
A press conference was held to announce plans to vent the Unit 1 PCV.
Workers attempted to enter the reactor building airlock door to perform surveys. As soon as the door was opened,
workers saw steam and closed the door. No surveys were performed.
Operators in the control room repeatedly reviewed and confirmed actions needed to vent containment. They
reviewed how to operate the valves, the sequence of operation, the physical location of the valves, and so forth.
Operators collected the equipment needed to perform the evolution, including fire-fighting turnout gear, SCBAs,
dosimeters, survey meters, and flashlights.
Exposure estimates associated with venting were provided to the government and associated authorities.
PCV pressure unexpectedly decreased and stabilized near 113 psi (0.78 MPa) without venting.

75Unit 1 Validated Event Timeline
Date

Time

12-Mar0423

12-Mar
12-Mar
12-Mar0430

12-Mar0500

12-Mar0514

12-Mar0544

12-Mar0546

12-Mar--

12-Mar--

12-Mar--

12-Mar
12-Mar0552

12-Mar

0633

0450
0450

0630

INPO 11-005

Unit 1 Actions
Increasing radiation levels were observed in the area around the main gate. Dose rates increased from 0.0069
mrem/hr (0.069 µSv/hr) at 0400 to 0.059 mrem/hr (0.59 µSv/hr) at 0423. This was reported to the government andother agencies at 0455.
The ERC informed the control room that field operations were prohibited because of tsunami warnings.
A dose rate of 0.1 mrem/hr (1 µSv/hr) is measured at the site boundary.
Workers were directed to wear full face masks with charcoal filters and coveralls when in the field.
Operators in the control room were directed to wear full face masks with charcoal filters and coveralls. Dose rates
in the control room increased to the point that operators moved to the Unit 2 side of the room.
Radiation dose rates increased in the plant concurrent with a decrease in drywell pressure. Workers determinedthese parameters indicated a drywell leak. This was reported to the government.
Radiation levels at the site boundary increased, and the Prime Minister expanded the evacuation zone around
Fukushima Daiichi to 6.2 miles (10 km).
A fire engine began injecting fresh water from the fire protection water storage tank into the reactor via the corespray system.
Initially, the fire engine was filled at the fire protection water storage tank, then driven close to the reactor building
to allow water to be injected through the fire protection line that was connected to the core spray line. The fire
protection water storage tank was at a low elevation, and the discharge pressure of the fire engine pump was
believed to be insufficient to overcome reactor pressure and inject into the core. Further complications, such as
having to drive under a half-collapsed building between the tank and the unit, further delayed injection.
After some trial and error, continuous water injection was commenced by connecting the fire engine to the tank,without the fire engine having to be at the low elevation of the tank. A hose was run from the suction of the fireengine to the fire protection water storage tank, and this allowed the fire engine to discharge water to the fire
protection line into the RPV.
An additional fire engine arrived on site and was used to transport fresh water from the fire protection tank at Unit
3 to the fire protection tank at Unit 1 repeatedly. The Unit 1 fire protection tank had only one hose connection, so
water injection into the RPV had to be stopped each time the second fire engine needed to replenish the water inthe tank.
A total of 264 gallons (1,000 liters) of fresh water was injected via the fire protection system.
A total of 528 gallons (2,000 liters) of fresh water was injected via the fire protection system.
TEPCO confirmed that some residents of Ookuma-machi, which is inside the evacuation zone, had not evacuatedyet. The residents had not left because they were not sure in which direction to evacuate.

76Unit 1 Validated Event Timeline
Date

Time

12-Mar0650

12-Mar
12-Mar0711

12-Mar0755

12-Mar
12-Mar
12-Mar0803

12-Mar0827

12-Mar0830

12-Mar0837

12-Mar0903

12-Mar--

12-Mar
12-Mar
12-Mar0904

12-Mar0915

12-Mar0924

12-Mar0930

12-Mar

0940

0755

0804
0815

0905
0915

INPO 11-005

Unit 1 Actions
The Minister of Economy, Trade and Industry ordered venting of units 1 and 2 containment in accordance with the
Act on Special Measures Concerning Nuclear Emergency Preparedness.
The Prime Minister arrived at the station.
A total of 793 gallons (3,000 liters) of fresh water was injected via the fire protection system.
The lowest indicated reactor water level decreased to the top of active fuel (TAF) (indicating level was higher thanactual level).
The site superintendent directed venting to start at 0900.
The Prime Minister left the station.
A total of 1,057 gallons (4,000 liters) of fresh water was injected via the fire protection system.
A TEPCO employee who had been dispatched to check the status of evacuations reported that the evacuation of
Ookuma-machi was not yet completed.
A total of 1,321 gallons (5,000 liters) of fresh water was injected via the fire protection system.
The Fukushima Prefectural government was informed that venting would start at approximately 0900. Venting was
being coordinated to ensure the evacuation was completed prior to venting commencing.
Evacuations south of the plant (Ookuma-machii) were confirmed as completed.
The control room operators formed three teams to perform the venting, with two operators on each team (one toperform actions and the other to assist by holding flashlights and monitoring dose rates and for other safetyconcerns, such as ongoing aftershocks). Because there was no means of communicating with the field teams, thedecision was made to dispatch one team at a time, with the next team leaving only after the preceding teamreturned.
The first team of operators was dispatched to open the motor-operated containment vent valve.
Venting of Unit 1 containment was announced to the press.
A total of 1,585 gallons (6,000 liters) of fresh water was injected via the fire protection system.
At approximately 0915, operators opened the motor-operated containment vent valve approximately 25 percent, as
directed by the procedure. The two operators involved each received a dose of 2.5 rem (25 mSv).
The second team of operators left the control room to open the torus vent valve.
The second team of operators was unsuccessful in the attempt to manually open the suppression chamber airoperated vent valve. The operators entered the torus room but had to turn back because they expected they would
exceed their 10 rem (100 mSv) dose limit.
A total of 5,548 gallons (21,000 liters) of fresh water was injected via the fire protection system.

77Unit 1 Validated Event Timeline
Date

Time

12-Mar--

12-Mar0953

12-Mar1017

12-Mar

1040

12-Mar1115

12-Mar--

12-Mar1139

12-Mar1205

12-Mar

1400

12-Mar

1430

12-Mar
12-Mar
12-Mar

1450
1453
1454

INPO 11-005

Unit 1 Actions
The control room operators decided not to dispatch the third team to the field because of the doses received andnotified the ERC of the inability to open the air-operated vent valve. The ERC then began working on methods toopen the suppression chamber air-operated vent valve. This would require DC power and a temporary air source.
The ERC also instructed the control room to attempt to operate the small air-operated valve remotely, hoping there
would be sufficient residual air pressure in the system to operate the valve.
TEPCO provided an evaluation of the potential exposure from attempting the venting operation to the government
and associated authorities.
Operators attempted to open the suppression chamber air-operated vent valve remotely with residual air pressure in
the instrumentation air system and temporary DC supplied by batteries. The control room operators attempted toopen the small air-operated valve three times (1017, 1023, and 1024).
Radiation levels increased at the main gate and monitoring posts. Workers assumed that this was probably fromventing the Unit 1 PCV.
Radiation levels were decreasing, which indicated that the venting was not fully effective. PCV pressure stabilized at
approximately 109 psia (0.75 MPa abs).
The ERC team was informed that a small air compressor was in a subcontractor’s office. Workers retrieved drawings
and took pictures of the connection point and planned how to install the compressor to enable remote operation of
the air-operated valve in the vent system from the control room.
The government was notified that one of the operators who had entered the torus room to attempt to vent the PCV
had received 10.6 rem (106 mSv) radiation dose.
Reactor water level decreased to 59 inches (1,500 mm) below TAF.
A temporary air compressor was located and transferred to the equipment bay, and needed fittings werelocated. The compressor was installed in the equipment bay to provide workers with some protection fromincreasing dose rates. At approximately 1400, the compressor was installed and started.
Operators opened the large suppression chamber air-operated vent valve. The venting evolution and release ofradioactive materials was confirmed based on an indicated decrease in containment pressure. It was reported to thegovernment at 1518.
Venting resulted in containment pressure lowering to 84 psia (0.58 MPa).
Fire engines had injected a total of 80 tons of fresh water into the reactor.
The site superintendent directed the injection of seawater into the reactor.

78

Unit 1 Validated Event Timeline
Date

Time

12-Mar--

12-Mar--

12-Mar1518

12-Marapx
1530

12-Mar1536

12-Mar1536

12-Mar1627

12-Mar1720

12-Mar1825

12-Mar1830

12-Mar1904

INPO 11-005

Unit 1 Actions
The Unit 1 fire protection tank was running out of water. Water transfers from other tanks continued, and work topump seawater into the reactors commenced.
Workers began using a main condenser backwash valve pit, which flooded during the tsunami, as a water source tosupply the fire pumps. The pit was closer to the units and at a higher elevation than the sea, making injection easier.
Three fire engines were connected in series to increase the discharge pressure.
The government was informed that the standby liquid control system was being recovered to allow borated water tobe injected into the reactor as soon as the system was operable. Seawater injection into the reactor through the fire
protection piping was also planned to start as soon as the lineup was competed and power was available.
Work to provide temporary power from a mobile generator to the Unit 2 standby liquid control system wascompleted.
A hydrogen explosion occurred in the reactor building (secondary containment).
The explosion caused extensive damage to the reactor building and injured five workers. Debris ejected by the
explosion damaged the temporary power cables, along with one of the large portable generators. The temporary
power supply for the standby liquid control system and the hoses that had been staged for seawater injection weredamaged beyond use. Although the fire engines were damaged, they were still usable. The injured workers werecarried to safety. Station workers, including the personnel working on the standby liquid control system and laying
temporary power cables, had to evacuate for an accountability. The area surrounding Unit 1 was strewn with highly
radioactive debris, so cleanup required support from radiation protection personnel.
Radiation dose rates at the monitoring post reached 101.5 mrem/hr (1,015 µSv/hr), which exceeded the 50 mrem/hr(500 µSv/hr) limit specified in Article 15, clause 1 of the Act on Special Measures Concerning Nuclear Emergency
Preparedness (abnormal increase in radiation dose at the site boundary). This was reported to the authorities.
Workers began inspecting the condition of fire trucks, buildings, and other facilities following the explosion.
The Prime Minister ordered the population within 12.4 mile (20 km) radius of Fukushima Daiichi Nuclear Power
Station to evacuate.
Field inspections revealed the area around the units was littered with debris, and the equipment that had been
staged to provide power to the standby liquid control system and hoses staged to inject seawater had been damaged
and were no longer usable.
The injection of nonborated seawater into the reactor commenced using the fire engines.

79Unit 1 Validated Event Timeline
Date

Time

12-Mar2045

12-Mar
14-Mar
19-Mar2045

20-Mar

1546

23-Mar--24-Mar25-Mar

1130

0110
--

1537

INPO 11-005

Unit 1 Actions
Operators commenced injecting water into the reactor through the core spray lines with borated seawater using fireengines.
Workers began adding boron to the seawater being injected into Unit 1.
All injection to the reactor was stopped because of a low level in the main condenser backwash valve pit.
A cable from a nearby transmission line was connected to the station.
A 480-volt emergency low-voltage switchboard (power center 2C) was energized using power supplied by atemporary line from the Tohoku nuclear station power line.
Reactor injection was changed to the feedwater system because of indications that the core spray injection nozzles
were clogging from salt accumulation.
Lighting was restored to the units 1-2 control room .
Reactor injection was changed from seawater to fresh water.

80Unit 2 Validated Event Timeline
8.2

INPO 11-005

Unit 2
Unit 2 Actions
Earthquake causes seismic motion in excess of the reactor protection system scram setpoints. A level 3 state of
emergency was automatically declared for the earthquake.
Automatic reactor scram signal on seismic trip
The main turbine automatically tripped.
Emergency diesel generators started and loaded.
Operators start reactor core isolation cooling (RCIC) to control reactor water level and pressure.
RCIC automatically tripped due to high reactor water level.
Operators confirmed the reactor was subcritical.
Operators manually started RCIC.
The Headquarters for Major Disaster Countermeasures was established in the TEPCO office in Tokyo to assess damages
from the earthquake and recover from the electric outage in TEPCO's service area.

Date

Time

11-Mar1446

11-Mar1447

11-Mar1447

11-Mar1447

11-Mar1450

11-Mar1451

11-Mar1501

11-Mar1502

11-Mar1506

11-Mar

1507

Torus cooling was placed in service.

11-Mar1525

11-Mar1527

11-Mar1528

11-Mar1535

11-Mar1539

11-Mar1541

Torus spray was placed in service.
The first wave of a series of tsunamis, generated by the earthquake, arrived at the station.
RCIC automatically tripped on high reactor water level.
The second tsunami hit the station.
Operators manually restarted RCIC.
A loss of all AC power occurs.

11-Mar--

11-Mar

1542

11-Mar--Lighting and indications were lost as AC and DC power systems failed. Control room lighting, including emergency
lighting, failed, leaving the control room completely dark.
TEPCO entered its emergency plan because of the loss of all AC power, in accordance with Article 10, paragraph 1 of the
Nuclear Disaster Law. Government offices were notified. The corporate Emergency Response Center was established.
Flooding caused a loss of the instrumentation power system, resulting in a major loss of control board instrumentation,
including reactor water level. The operating status of RCIC was unknown because of the control board instrumentation
loss.

81

Unit 2 Validated Event Timeline
Date

Time

11-Mar--

11-Mar--

11-Mar--

11-Mar--

11-Mar1636

11-Mar--

11-Mar1712

11-Mar--

11-Mar--

11-Mar2049

INPO 11-005

Unit 2 Actions
Inspections revealed that switchyard breaker damage would prevent off-site power from being restored soon. Diesel
generators were not operable and restoration was difficult because either the engines or the associated switchgear were
submerged. It was concluded that power restoration using power supply vehicles was necessary.
Although high- and low-voltage power supply vehicles headed to Fukushima Daiichi, they were not able to drive to thesite quickly because of traffic and damaged roads. TEPCO considered transporting the power supply vehicles using
helicopters, but the power supply vehicles weighed too much. TEPCO requested that the Tohoku Electric Power Company
dispatch power supply vehicles to Fukushima Daiichi.
Visual inspections and insulation resistance tests were performed on metal clad switchgear (M/C) and power centers(P/Cs) to determine the extent of the damage from the tsunami. All of the Unit 1 and Unit 3 M/C and P/Cs were
inoperable. At Unit 2, all of the M/C and some of the P/Cs at were inoperable. Later, it was confirmed four of seven P/Cs
were operable.
TEPCO decided to connect a power supply vehicle to the primary side of a Unit 2 power transformer (6.9-kV/480V) to
provide 480 volts to energize the control rod drive and standby liquid control pumps. Considering the distance to the P/Cof Unit 2 and the ability to run cables, the power supply vehicle was driven to the side of the Unit 2 turbine building.
TEPCO again entered the emergency plan in accordance with Article 15, paragraph 1 of the Nuclear Disaster Law because
operators concluded they could not determine reactor water level and the status of injection into the reactor. The
government was notified at 1645.
Temporary batteries and cables were gathered and carried to the units 1-2 control room. After confirming the wiring
layout using drawings, batteries were connected to instrument panels. The top priority was to verify the status of water
injection into the RPV, so efforts were focused on connecting batteries to the DC-powered reactor water level indicator.
The site superintendent directed workers to investigate methods of injecting water into the reactor using the fire
protection system or fire trucks.
Operators reviewed methods for alternative water injection using the accident management (AMG) procedures andconfirmed the injection path using the installed fire pump into the RPV via the core spray system.
Operators established an alternative water injection flow path via the residual heat removal system by manually opening
valves in the turbine and reactor building to establish an injection path after the reactor pressure decreased to 100 psi(0.69 MPa). This was done before the reactor was depressurized so the valves would be in the proper position shoulddose rates increase and access become difficult.
A small portable electric generator was installed, which restored some temporary lighting in the units 1-2 control room.

82

Unit 2 Validated Event Timeline

INPO 11-005

Date

Time

Unit 2 Actions

11-Mar2050

11-Mar2102

11-Mar2113

The Fukushima Prefecture ordered evacuation of the population within 1.2 mile (2 km) radius of the station.
Operators did not have reactor water level indication in the control room and were unable to verify the injection of water
into the reactor using RCIC. The operators notified the government of the potential for water level to lower to the top of
active fuel (TAF).
The government was notified that reactor water level was estimated to reach TAF at 2140.

11-Mar2123

11-Mar2150

11-Mar--

11-Mar--

11-Mar2325

12-Mar--

12-Mar0030

12-Mar--

12-Mar0130

12-Mar0145

The Prime Minister ordered the evacuation of inhabitants within a 1.9 mile (3 km) radius of the station and sheltering for
those within a 1.9 to 6.2 mile (3 to 10 km) radius.
Water level indication was restored in the control room; level indicated 134 inches (3,400 mm) above TAF.
Late in the evening of March 11 and into the early morning of March 12, power supply vehicles from Tohoku ElectricPower Company and TEPCO arrived at the site.
TEPCO's power supply vehicles were given priority for connecting cables.
Drywell pressure indication was restored, and indicated pressure was 20 psia (0.141 MPa).
Operators reviewed valve drawings to determine if they could open the valves required for venting manually, or whether
they would need to attach some type of actuator to force the valve open. Based on their reviews of the piping andinstrumentation diagrams, accident management procedures, and the vent operating procedure, the operatorsdeveloped a method for venting Unit 2. The motor-operated valve could be opened manually, but the suppression
chamber air-operated vent valve could not. The operators prepared a venting procedure and reviewed the locations of
the vent valves.
The government confirmed the evacuation of the population within 1.9 miles (3 km) was completed.
Wearing breathing protection and boots, workers attempted to verify the condition of RCIC, but field conditions madethis very difficult. Work that would normally take approximately 10 minutes required more than one hour to complete.
At around 0100, operators went to the RCIC room. The room was dark, and the water level in the room came nearly tothe top of the workers’ boots , so they turned back without getting close enough to check system operation. However,they could hear faint metallic sounds, indicating the system was operating. Because of the lack of communications
methods, the workers had to return to the control room to report their findings.
At approximately 0130, TEPCO officials informed the Prime Minister, Minister of METI, and NISA of their proposal to ventthe containment. Concurrence was received from the agencies.
The government confirmed the evacuation of the population within 1.9 miles (3 km) was completed.

83Unit 2 Validated Event Timeline
Date

Time

12-Mar0200

12-Mar0255

12-Mar0306

12-Mar0333

12-Mar0423

12-Marapx.
0500

12-Mar0544

12-Mar0650

12-Mar0711

12-Mar0804

12-Mar--

12-Mar--

12-Mar--INPO 11-005

Unit 2 Actions
At around 0200, workers made another attempt to verify RCIC operation. This time, water flooding in the RCIC room had
increased, and workers could not enter the room. On this attempt, workers checked reactor pressure and RCIC pumpdischarge pressure on an instrument rack in the reactor building. RCIC pump discharge pressure was high, so workers
concluded RCIC was operating.
Field workers returned to the control room to report that RCIC was in operation on Unit 2. The control room operatorsforwarded this information to the Emergency Response Center. Based on this report, venting of the Unit 1 containment
was given a higher priority. The operators proceeded with the lineup to vent Unit 1 and, at the same time, monitored
parameters on Unit 2.
A press conference was held to announce plans to vent the PCV.
An evaluation of the potential off-site exposure from venting was provided to the government.
Increasing radiation levels were observed in the area around the main gate. Dose rates increased from 0.0069 mrem/hr(0.069 µSv/hr) at 0400 to 0.059 mrem/hr (0.59 µSv/hr) at 0423. The government and associated authorities were notified
at 0455.
Operators in the control room were directed to wear full face masks with charcoal filters and coveralls. Dose rates in thecontrol room increased to the point that Unit 1 operators moved to the Unit 2 side of the control room.
The Prime Minister ordered evacuation of the population out to 6.2 miles (10 km) from the station.
The Minister of Economy, Trade and Industry ordered venting of units 1 and 2 containment in accordance with the Act on
Special Measures Concerning Nuclear Emergency Preparedness.
The Prime Minister arrived at Fukushima Daiichi Nuclear Power Station.
The Prime Minister departed Fukushima Daiichi Nuclear Power Station.
The cable was more than 4 inches (10 cm) in diameter; about 656 feet (200 meters) long, and weighed more than 1 ton.
Laying this kind of cable usually takes days using machinery, but 40 TEPCO personnel laid the cable in four to five hours byhand. It was very difficult work because of extremely bad working conditions, such as darkness, flooding from thetsunami, debris, and lost manhole covers. It was challenging to locate cable penetrations in the dark, and some doors
needed to be destroyed to route the cable. Earthquake aftershocks and recurring tsunami warnings caused work to beinterrupted periodically.
Work to connect the cable to the power center was performed by several engineers, and the job took several hours.
The lack of power challenged communications, causing workers to have to walk to communicate, which was very timeconsuming.

84

Unit 2 Validated Event Timeline
Date

Time

12-Mar1530

12-Mar1536

12-Mar1627

12-Mar1730

12-Mar--

12-Mar1825

13-Mar0810

13-Mar0856

13-Mar13-Mar

1015
1035

13-Mar

1100

13-Mar1120

13-Mar1205

INPO 11-005

Unit 2 Actions
The primary side of the power cabinet was connected to the temporary cable, completing the lineup for temporary power
to standby liquid control.
An explosion occurred inside the Unit 1 reactor building, resulting in a large amount of flying debris. The temporary
cables, generators, and hoses that had been staged for injecting seawater were damaged. Workers had to evacuate to
the Emergency Response Center.
The radiation dose rate at the monitoring post reached 101.5 mrem/hr (1.015 mSv/hr), which exceeded the 50 mrem/hr(500 µSv/hr) limit specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal increase inradiation dose at the site boundary. This was reported to the authorities.
The site superintendent ordered preparations to begin for venting the containment.
RCIC injection continued, and drywell pressure was stable at 29–44 psia (200-300 kPa abs.). Knowing that venting would
be required eventually, operators commenced preparations for lining up to vent the Unit 2 and Unit 3 containments.
Operators planned to manually open the vent valves (other than the rupture disk) because the dose in the field wascurrently low.
The Prime Minister ordered the evacuation of the population within a 12.4 mile (20 km) radius of Fukushima Daiichi.
The shift supervisor, wearing a self-contained breathing apparatus (SCBA) and protective equipment, opened the motoroperated containment vent valve 25 percent in accordance with the procedure. Containment pressure was 50.8 psia
(0.35 MPa abs).
The dose rate measured 88.2 mrem/hr (882 µSv/h) at the monitoring post, exceeding the 50 mrem/hr (500 µSv/hr) limit
specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal increase in radiation dose at thesite boundary. This was reported to the authorities at 0901.
The site superintendent ordered operators to vent the containment.
Drywell pressure indicated 1.45 psia (0.010 MPa abs).
Using the small generator installed for the temporary lighting in the control room as a power source, operators opened
the large suppression chamber air-operated vent valve. The vent lineup was completed, but the rupture disk remained
intact because drywell pressure remained below the 62 psig (427 kPa gauge) rupture pressure of the disk.
A press release was made to announce venting the Unit 2 containment.
The site superintendent ordered preparations for injecting seawater into the reactor.

85

Unit 2 Validated Event Timeline
Date

Time

13-Mar1415

13-Mar1518

14-Mar0220

14-Mar0240

14-Mar0400

14-Mar0912

14-Mar1101

14-Mar--

14-Mar--

14-Mar--

14-Mar1130

14-Mar--INPO 11-005

Unit 2 Actions
The dose rate measured 90.5 mrem/hr (905 µSv/h) at the monitoring post, exceeding the 50 mrem/hr (500 µSv/hr) limit
specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal increase in radiation dose at thesite boundary. This was reported to the authorities at 1423.
The projected off-site dose associated with venting was reported to the authorities.
The dose rate measured 75.1 mrem/hr (751 µSv/hr) at the monitoring post, exceeding the 50 mrem/hr (500 µSv/hr) limit
specified in Article 15, paragraph 1 of Nuclear Disaster Law, indicating an abnormal increase in radiation dose at the siteboundary. This was reported to the authorities at 0424.
The dose rate measured 65 mrem/hr (650 µSv/hr) at the monitoring post, exceeding the 50 mrem/hr (500 µSv/hr) limit
specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal increase in radiation dose at thesite boundary. This was reported to the authorities at 0537.
The dose rate measured 82 mrem/hr (820 µSv/hr) at the monitoring post, exceeding the 50 mrem/hr (500 µSv/hr) limit
specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal increase in radiation dose at thesite boundary. This was reported to the authorities at 0800.
The dose rate measured 51.9 mrem/hr (518.7 µSv/hr) at the monitoring post, exceeding the 50 mrem/hr (500 µSv/hr)
limit specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal increase in radiation dose at
the site boundary. This was reported to the authorities at 0934.
The Unit 3 reactor building exploded.
The explosion damaged the fire engine and hoses that had been staged to inject seawater into the reactor.
The explosion knocked the power supply loose from the solenoid for the control air valve to the suppression chamberlarge air-operated vent valve, and the valve closed. The venting lineup had to be performed again.
After the explosion, all workers except the shift workers in the control room were evacuated to the Emergency Response
Center. The status of the work in the field had to be confirmed and resumption of work was delayed.
Containment pressure was stable around 65.3 psia (450 kPa abs) to 66.7 psia (460 kPa abs), stable below the pressure
needed to open the rupture disk.
Shortly after 1200, a site survey confirmed scattered contaminated rubble and equipment damage caused by the Unit 3
explosion. A decision was made to change the water supply and draw a suction directly from the shallow draft quay
(harbor) instead of the main condenser backwash valve pit of Unit 3, which was now covered with contaminated debris.
In the high radiation environment surrounded by scattered rubble, personnel prepared a new water injection line byusing available fire engines and hose pipes.

86

Unit 2 Validated Event Timeline
Date

Time

14-Mar1230

14-Mar1305

14-Mar1318

14-Mar1325

14-Mar1443

14-Mar--

14-Mar1528

14-Mar1600

14-Mar--

14-Mar1600

14-Mar1600

14-Mar1620

14-Mar
14-Mar

1620
1630

INPO 11-005

Unit 2 Actions
Operators were concerned about the ability to depressurize the reactor with the suppression chamber saturated.
Suppression chamber temperature was 301oF (149.3oC) and pressure was 70.5 psia (486 kPa abs). Operators decided tovent containment first, then open a safety relief valve (SRV) and start water injection.
Workers realigned fire hoses and fire engines in preparation for injecting seawater into the reactor.
Reactor water level was 94.5 inches (2400 mm) above TAF and trending downward. The government and otherauthorities were notified of the ongoing preparations to inject seawater.
Based on a lowering trend in reactor water level, operators concluded that RCIC may have failed. TEPCO determined thisconstituted a loss of reactor cooling in accordance with Article 15, paragraph 1 of the Nuclear Disaster Law. The
government and associated authorities were notified at 1338. At the time of the trip, indicated reactor water level wasapproximately 94.5 inches (2,400 mm) above the top of active fuel and drywell pressure was 67 psi (465 kPa).
A fire engine was connected to the fire protection system.
Between 1500 and 1600, frequent aftershocks made work difficult as workers repeatedly had to take refuge.
The government and associated authorities were notified that reactor water level was estimated to reach TAF at 1630.
Because of delays in opening the suppression chamber vent valve, the priority for reactor depressurization was changed.
The site superintendent directed that the work to open the suppression chamber vent valve be performed in parallel withreactor depressurization.
Because of a lack of power, temporary batteries were necessary to open the SRV. Batteries were gathered from cars,
carried to the control room, and connected. However, the voltage was not adequate, so additional batteries were
scavenged and added. Operators attempted to operate several SRVs. Efforts to depressurize the reactor continued until
around 1800, when depressurization began. The high suppression chamber temperature and pressure resulted in thereactor depressurizing more slowly than desired.
After receiving permission to recommence work after the explosion, operators continue attempts to open the large
suppression chamber air-operated vent valve.
Reactor water level decreased to 12 inches (300 mm) above TAF.
The large suppression chamber air-operated vent valve could not be opened because of insufficient air pressure from theair compressor.
Indicated reactor water level reached TAF.
The fire engine was started in preparation for injecting seawater to the reactor.

87Unit 2 Validated Event Timeline
Date

Time

14-Mar1634

14-Mar1717

14-Mar

14-Mar

1800
apx.
1800
1803

14-Mar

1822

14-Mar

14-Mar--

14-Mar1903

14-Mar
14-Mar

1920

14-Mar1954

14-Mar

1957

14-Marapx.
2100

14-Mar2120

14-Mar2130

14-Mar2135

INPO 11-005

Unit 2 Actions
The government and associated authorities were notified that in addition to the start of reactor depressurization,
seawater injection would commence using the fire protection system. Reactor pressure was 1,015 psi (6.998 MPa).
Reactor water level decreased to TAF. This was reported to the authorities at 1725.
Reactor pressure was 1,035 psig (7.133 MPa gauge).
At approximately 1800, operators opened an SRV and began depressurizing the reactor.
Reactor pressure was 881 psig (6.075 MPa gauge).
Reactor water level lowered to 146 inches (3,700 mm) below TAF, the bottom of the indicating range, indicating the core
was completely uncovered. This was reported to the authorities at 1932.
Despite attempts to vent containment, pressure remained stable. Operators began work to open the small suppression
chamber air-operated vent valve in addition to the large suppression chamber air-operated vent valve to reopen the ventpath. However, the large suppression chamber air-operated vent valve would not open. Workers believed the solenoidon the control air valve had malfunctioned.
Reactor pressure stabilizes at approximately 91.4 psig (0.63 MPa gauge).
The radiation on site was high, forcing workers to make rounds to check the operating status of the fire engines.
While touring to check the status of the fire engines, workers discovered that an engine had run out of fuel and noseawater was being injected into the reactor.
After a fire engine was refueled and started, seawater injection commenced into the reactor via the fire protection
system.
A second fire engine was started to augment seawater injection into the reactor.
Operators opened the small suppression chamber air-operated vent valve. This established the venting lineup except therupture disk. Containment pressure was lower than the working pressure of the rupture disk at 62 psig (427 kPa gauge),
so venting did not occur. The vent valves remained open and operators continued to monitor containment pressure.
Indicated containment pressure was 60.8 psia (0.419 MPa abs).
A second SRV was opened, and reactor water level trended up. This was reported to the government and associatedauthorities at 2134.
Reactor water level was 118 inches (3,000 mm) below TAF.
The dose rate measured by a monitoring car was 76 mrem/hr (760 µSv/hr), exceeding the 50 mrem/hr (500 µSv/hr) limit
specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal increase in radiation dose at thesite boundary. This was reported to the authorities at 2235.

88

Unit 2 Validated Event Timeline
Date

Time

14-Mar

2200

14-Mar2250

14-Mar2250

14-Mar2325

14-Mar2330

14-Mar--

14-Mar--15-Mar0002

15-Mar0128

15-Mar

0300

15-Mar0600

15-Mar0650

INPO 11-005

Unit 2 Actions
Reactor water level recovered to 63 inches (1,600 mm) below TAF, trending up.
Containment pressure increased to 78 psia (0.54 MPa abs), which exceeded the maximum design pressure of 62 psig (427kPa gauge). This represented an abnormal increase in containment pressure as specified in Article15, paragraph 1 of the
Nuclear Disaster Law. This was reported to the authorities at 2339.
Reactor pressure increased unexpectedly, accompanied by a drop in reactor water level. Reactor pressure increased from62 psig (0.428 MPa gauge) at 2240 to 264.4 psig (1.823 MPa gauge) at 2250. At the same times, reactor water level
lowered from 27.6 inches (700 mm) below TAF to 63 inches (1,600 mm) below TAF.
Containment pressure increased above 101.5 psia (0.7 MPa abs).
Reactor pressure again began to drop. The highest reactor pressure was 456.9 psig (3.15 MPa gauge).
Drywell pressure was trending up and reached 101.5 psia (0.7 MPa abs); but in contrast, suppression chamber pressure
remained stable around 43.5 to 58 psia (300 to 400 kPa abs). The pressure was not equalizing between the containment
and the suppression chamber.
Pressure in the suppression chamber was lower than rupture disk pressure and drywell pressure was increasing, so
operators decided to open the small drywell air-operated vent valve.
Operators opened the small drywell air-operated vent valve. The vent line lineup was completed to vent from thedrywell, except for the rupture disk, which remained closed. Containment pressure remained stable at approximately 109
psia (750 kPa abs). A few minutes later, operators checked and noted that the small air-operated valve had closed.
Reactor pressure decreased to 91 psig (0.63 MPa gauge).
Containment pressure exceeded maximum design pressure, and operators again attempted to vent the containment to
reduce pressure but were unsuccessful. This was reported to the government and associated authorities at 0417.
At around 0600, a loud noise was heard in the area around the torus. While suppression chamber pressure dropped to 0
psia (0 Mpa abs), indicating a potential instrument failure, drywell pressure remained high, indicating 105.9 psia (0.73
MPa abs). At that time, reactor water level was 110 inches (2,800 mm) below TAF.
The radiation dose at the main gate was 58.4 mrem/hr (583.7 µSv/hr), exceeding the 50 mrem/hr (500 µSv/hr) limit
specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal increase in radiation dose at thesite boundary. This was reported to the authorities at 0700.

89Unit 2 Validated Event Timeline
Date

Time

15-Mar

0700

15-Mar0811

15-Mar0825

15-Mar

1100

15-Mar1125

15-Mar1600

15-Mar2305

INPO 11-005

Unit 2 Actions
The government was notified that all nonessential personnel (650 people) were evacuated to Fukushima Daini, leaving 70
people at the station. The shift operators periodically went to the control room to log containment pressure and othercritical parameters.
The radiation dose at the main gate was 80.7 mrem/hr (807 µSv/hr), exceeding the 50 mrem/hr (500 µSv/hr) limit
specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal emission of radioactive materials
from a fire or explosion. This was reported to the authorities at 0836.
Workers observed steam rising from the area around the fifth floor of the reactor building. This was reported to theauthorities at 0918.
The Prime Minister directed that residents living between 12.4 and 18.6 miles (20 and 30 km) of the station take shelter.
Containment pressure decreased from 106 psia (730 kPa abs) at 0720shortly after a loud noise was heard around thetorusto 22.5 psia (155 kPa abs) at 1125. No information is available regarding containment pressure between the twotimes.
The radiation dose at the main gate was 53.2 mrem/hr (531.6 µSv/hr), exceeding the 50 mrem/hr (500 µSv/hr) limit
specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal increase in radiation dose at thesite boundary. This was reported to the authorities at 1622.
The radiation dose at the main gate was 454.8 mrem/hr (4548 µSv/hr), exceeding the 50 mrem/hr (500 µSv/hr) limit
specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal increase in radiation dose at thesite boundary. This was reported to the authorities at 2320.

90Unit 3 Validated Event Timeline
8.3

Unit 3Date

Time

11-Mar1446

11-Mar1447

11-Mar

11-Mar

1447
apx.
1448
1451

11-Mar1454

11-Mar1505

11-Mar1506

11-Mar1515

11-Mar1525

11-Mar1527

11-Mar1535

11-Mar--

11-Mar

1538

11-Mar

1542

11-Mar1603

11-Mar1636

11-Mar2050

11-Mar2123

11-Mar2158

12-Mar0030

12-Mar0145

11-MarINPO 11-005

Unit 3 Actions
An earthquake caused seismic motion in excess of the reactor protection system scram setpoints. A level 3 state of
emergency was automatically declared for the earthquake.
Automatic reactor scram signal on seismic trip
Operators manually tripped the main turbine.
Emergency diesel generators started and loaded.
Operators commenced breaking main condenser vacuum.
Operators confirmed the reactor was subcritical.
The operators manually initiated reactor core isolation cooling (RCIC).
The Headquarters for Major Disaster Countermeasures was established in the TEPCO office in Tokyo to assess damages
from the earthquake and recover from the electric outage in TEPCO's service area.
Operators completed breaking main condenser vacuum.
RCIC tripped on high reactor water level.
The first wave of a series of tsunamis, generated by the earthquake, arrived at the station.
The second tsunami hit the station.
Unnecessary DC loads were shed to extend battery life.
All AC power was lost.
TEPCO entered its emergency plan because of the loss of all AC power, in accordance with Article 10, paragraph 1 of the
Nuclear Disaster Law. Government offices were notified. The corporate Emergency Response Center was established.
Operators manually restarted RCIC and prepared HPCI for startup.
The second emergency level was announced.
Fukushima Prefecture ordered evacuation of the population within a 1.2 mile (2 km) radius of the station.
The Prime Minister ordered the evacuation of the population within a 1.9 mile (3 km) radius of the station and sheltering
for the population within a 1.9 to 6.2 mile (3 km to 10 km) radius of the station.
A small portable electric generator was installed, which restored some temporary lighting in the units 3-4 control room.
The government confirmed the evacuation of the population within 1.9 miles (3 km) was completed.
The government reconfirmed the evacuation of the population within 1.9 miles (3 km) was completed.

91Unit 3 Validated Event Timeline
Date

Time

12-Mar0423

12-Mar0544

12-Mar0711

12-Mar0804

12-Mar1136

12-Mar--

12-Mar1210

12-Mar1235

12-Mar1730

12-Mar--

12-Mar1825

12-Mar--13-Mar0242

13-Mar--INPO 11-005

Unit 3 Actions
Increasing radiation levels were observed in the area around the main gate. Dose rates increased from 0.0069 mrem/hr(0.069 µSv/hr) at 0400 to 0.059 mrem/hr (0.59 µSv/hr) at 0423. The government and associated authorities were
notified at 0455.
The Prime Minister ordered the evacuation of the population out to a 6.2 mile (10 km) radius.
The Prime Minister arrived.
The Prime Minister departed.
RCIC tripped.
Station fire engines were being used to inject water into Unit 1. Additional fire engines were requested but were notavailable.
Primary containment pressure had slowly increased, peaking at 57 psia (0.39 MPa abs).
HPCI automatically started on low reactor water level.
The site superintendent ordered preparations for venting the containment to proceed.
HPCI injection continued. Knowing that venting would be required eventually, operators commenced preparations forlining up to vent the Unit 2 and Unit 3 containments. Operators planned to manually open the vent valves (other than
the rupture disk) because the dose in the field was currently low.
The Prime Minister ordered the evacuation of the population within a 12.4 mile (20 km) radius of the station.
After 2100, operators started a review of the vent procedures to identify the sequence and location of vent valves,
which were written on a whiteboard. Emergency response workers reviewed the vent procedure for Unit 1. Based onthat review and the accident management procedure for Unit 3, they developed a procedure to manually vent Unit 3
and provided it to the control room operators.
The HPCI system automatically tripped, which resulted in a loss of all water injection sources.
Operators attempted to restart water injection by installed equipment but were unsuccessful. HPCI could not be started
because the batteries were exhausted. Operators attempted to start RCIC locally, but it would not start. They next
attempted water injection by diesel-driven fire pump but could not because reactor pressure had increased as high as
approximately 580 psi (4 MPa). Attempts to restore power to the standby liquid control system were unsuccessful
because the poor working conditions (darkness, debris, flooding) prevented temporary power installation from being
completed.

92

Unit 3 Validated Event Timeline
Date

Time

13-Mar--13-Marapx.
0450

13-Mar0500

13-Mar0500

13-Mar0510

13-Mar0515

13-Mar0523

13-Mar0550

13-Mar0619

13-Mar0735

13-Mar0739

13-Mar0745

13-Mar0835

13-Mar0841

13-Mar0856

INPO 11-005

Unit 3 Actions
To inject water by fire engines, it was necessary to decrease the reactor pressure by operating SRVs, but no SRVs wereavailable because the batteries were depleted. All of the available batteries had already been used, so workers
scavenged batteries from cars and connected them to instrumentation in the control room and used them to power an
SRV.
To open the large suppression chamber air-operated vent valve, workers used the small generator that was installed toprovide temporary lighting in the control room to provide power to the valve solenoid. An operator checked the valve
indication locally at the torus room, but the valve indicated closed. The torus room was very hot and the room wascompletely dark, which made for a difficult working environment.
Reactor pressure increased to greater than 1,070 psi (7.38 MPa).
Reactor water level was 79 inches (2000 mm) below TAF and lowering.
Because the RCIC could not be restarted, TEPCO determined that a loss of reactor cooling had occurred in accordancewith Article 15, paragraph 1 of the Nuclear Disaster Law. Government offices were notified at 0558.
The site superintendent instructed operators to complete the lineup for venting the containment.
The solenoid for the large suppression chamber air-operated vent valve was energized, but the vent valve remained
closed. Operators determined that there was insufficient air pressure to operate the valve. The operators replaced atemporary air bottle that had been installed to provide air pressure, and the valve opened.
A press release was made to announce venting.
TEPCO determined reactor water level reached the top of active fuel (TAF) at 0415, and a notification was made to thegovernment offices.
The government offices were notified of the dose expected to be received during venting.
Containment spray was initiated in the containment using a fire engine. The government was notified at 0756.
Reactor water level decreased to 118 inches (3,000 mm) below TAF, reactor pressure was 1,060 psig (7.310 Mpa gauge),
and primary containment pressure increased to 67 psia (0.46 MPa abs).
Operators opened the motor-operated containment vent valve 15 percent, as directed by the procedure.
The operators opened the large air-operated torus vent valve, completing the vent lineup except the rupture disk. The
government was notified at 0846.
The dose rate measured 88.2 mrem/hr (882 µSv/h) at the monitoring post, exceeding the 50 mrem/hr (500 µSv/hr) limit
specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal increase in radiation dose at thesite boundary. This was reported to the authorities at 0901.

93Unit 3 Validated Event Timeline
Date

Time

13-Mar0908

13-Mar0924

13-Mar0925

13-Marapx.
0928

13-Mar0936

13-Mar1030

13-Mar1117

13-Mar--13-Mar1220

13-Mar1230

13-Mar1300

13-Mar1312

13-Mar--13-Mar--INPO 11-005

Unit 3 Actions
Workers had scavenged enough batteries to power the SRVs, and the operators opened an SRV to depressurize thereactor. At 0920, the government was notified that water injection would be started through a fire protection line.
Containment pressure lowered from 92 psia (0.637 MPa abs) at 0910 to 78 psia (0.540 MPa abs). The Emergency
Response Center determined that venting had successfully started at about 0920.
Reactor pressure lowered to 50.8 psig (0.35 MPa gauge), and injection of borated fresh water into the reactor using thefire system began.
The air cylinder providing air to the large air-operated vent valve was losing pressure. Workers investigated and found aleak on one fitting, which they repaired.
Operators verified that containment pressure was decreasing and that the injection of water into the reactor had
started. The government was notified.
The site superintendent ordered workers to be ready to commence injection using seawater.
The air-operated torus vent valve was found closed. The pressure in the air cylinder being used to provide air was
decreasing.
Attempts to lock open the valve were not successful because of the extreme adverse conditions in the torus room. The
room was dark and hot, and the torus was shaking because of the open SRV.
Injection of fresh water was stopped as the water ran out. Workers had prepared to make a swift change to seawater
injection, but an aftershock and subsequent evacuation order occurred while the work was being carried out, so theswap was delayed. Work was also ongoing to replenish the fresh water supply.
Workers replaced the air bottle, and the large air-operated torus vent valve was opened.
Reactor pressure decreased to 28 psi (0.19 MPa), and reactor water level was approximately 79 inches (2,000 mm)
below TAF by the lowest indication.
Workers commenced injecting seawater into the reactor using fire trucks and the fire protection system.
Requests for off-site fire engines were made, but it was difficult for those fire engines to reach the plant, because theroads were impassable and dose rates on the site were increasing. When they finally were able to come, it took longer
for the fire engines to come on station because the drivers needed to be changed to site workers off site.
Other sources for seawater were considered, including accumulated water in the basement of the Unit 4 turbine
building, which was accessed by breaking the entrance shutters of truck bay doors and moving fire engines into place.
However, this was unsuccessful. Workers also considered taking water from the discharge channel of Unit 4 or the
training center pool, but this also did not work.

94Unit 3 Validated Event Timeline

INPO 11-005

Date

Time

Unit 3 Actions

13-Mar1415

The dose rate measured 90.5 mrem/hr (905 µSv/h) at the monitoring post, exceeding the 50 mrem/hr (500 µSv/hr) limit
specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal increase in radiation dose at thesite boundary. This was reported to the authorities at 1423.

13-Mar1431

13-Mar

1500

13-Mar1528

13-Mar1752

13-Mar2010

14-Mar0110

14-Mar0200

14-Mar0220

14-Mar0240

14-Mar

0300

14-Mar

0310

14-Mar

0320

Workers moved the fire engine around, allowing the hose to drop deeper into the valve pit; and seawater injection into
the reactor was restored using a fire engine.

14-Mar0340

The solenoid valve for the small suppression chamber air-operated valve was energized in an attempt to lower PCV
pressure. Containment pressure was approximately 47.9 psia (0.33 MPa abs) and continued to trend up.

Radiation dose rates over 30 rem/hr (300 mSv/hr) were measured at the north side reactor building air lock, with visible
steam and 10 rem/hr (100 mSv/hr) at the south side air lock.
Containment pressure began to increase from 11.6 psia (0.08 MPa abs) to 13 psia (0.09 MPa abs).
The dose rate in the Unit 3 side of the control room reached 1.2 rem/hr (12 mSv/hr), and operators moved to the Unit 4
side.
A temporary compressor became available for use. Workers then began to attach the compressor to the instrument air
system to keep the suppression chamber air-operated vent valve open.
A reduction in containment pressure indicated that the suppression chamber air-operated vent valve was open. Overthe next five days, losses of air pressure or DC power to the solenoid caused the valve to periodically close.
Injection into the reactor was stopped because of a low level in the main condenser backwash valve pit being used as awater source.
Containment pressure was trending upward at 38.4 psia (0.265 MPa abs). Personnel decided to open the small
suppression chamber air-operated vent valve.
The dose rate measured 75.1 mrem/hr (751 µSv/hr) at the monitoring post, exceeding the 50 mrem/hr (500 µSv/hr)
limit specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal increase in radiation dose at
the site boundary. This was reported to the authorities at 0424.
The dose rate measured 65 mrem/hr (650 µSv/hr) at the monitoring post, exceeding the 50 mrem/hr (500 µSv/hr) limit
specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal increase in radiation dose at thesite boundary. This was reported to the authorities at 0537.
Containment pressure increased from 38.4 psia (0.265 MPa abs) at 0200, to 45.7 psia (0.315 MPa abs) at 0300.
Reactor water level decreased to 91 inches (2,300 mm) below TAF by the lowest indication.

95

Unit 3 Validated Event Timeline

INPO 11-005

Date

Time

Unit 3 Actions

14-Mar

0400

The dose rate measured 82 mrem/hr (820 µSv/hr) at the monitoring post, exceeding the 50 mrem/hr (500 µSv/hr) limit
specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal increase in radiation dose at thesite boundary. This was reported to the authorities at 0800.

14-Mar0520

14-Mar0610

14-Mar0610

14-Mar0912

14-Mar0920

14-Mar1053

14-Mar1101

14-Mar1101

14-Mar1200

14-Mar1630

14-Mar1640

14-Mar2230

Work began to open the small air-operated valve for the suppression chamber. By 0530, containment pressure was 56.6
psia (0.39 MPa abs) and increasing.
The small air-operated suppression chamber vent valve was confirmed open.
After confirming that the small suppression chamber air-operated valve was open, the operators could not maintain thevalve open because of a loss of air pressure and a loss of DC power to the solenoid.
The dose rate measured 51.9 mrem/hr (518.7 µSv/hr) at the monitoring post, exceeding the 50 mrem/hr (500 µSv/hr)
limit specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal increase in radiation dose at
the site boundary. This was reported to the authorities at 0934.
Additional fire engines arrived around sunrise, and workers began using two fire engines that had arrived from off site topump seawater from the shallow draft quay (harbor) into the main condenser backwash valve pit.
Seven 5-ton Self-Defense Force water supply vehicles arrived at the site and started to supply fresh water to the maincondenser backwash valve pit.
The Unit 3 reactor building exploded. The explosion damaged fire trucks and hoses. Seawater injection into the reactor
was lost.
All workers except shift personnel in the control room stopped work and evacuated to the Emergency Response Center.
Four TEPCO employees, 3 contractors, and 4 self-defense personnel were injured (11 total). The explosion scatteredhighly radioactive debris around the site. Because of the debris and radiation levels, the backwash valve pit was nolonger usable as a water source.
Containment pressure began to increase from 52.2 psia (0.36 MPa abs) to 53.7 psia (0.37 MPa abs).
The undamaged fire engine was moved to the shallow draft quay, and a new injection lineup was completed. Fireengines and hoses were rearranged to inject seawater directly from the shallow draft quay into the reactor. Connecting
two fire engines in series allowed seawater to supply both units 2 and 3.
Containment pressure began to lower from 69.6 psia (0.48 MPa abs).
Containment pressure began to increase from 52.2 psia (0.36 MPa abs), and reactor pressure vessel pressure was 26.1
psi (0.18 MPa) and increasing.

96

Unit 3 Validated Event Timeline

Date

Time

14-Mar2135

15-Mar0650

15-Mar

0700

15-Mar0755

15-Mar0811

15-Mar

1100

15-Mar1600

15-Mar15-Mar1600
1605

15-Mar2305

16-Mar0155

INPO 11-005

Unit 3 Actions
The dose rate measured by a monitoring car is 76 mrem/hr (760 µSv/hr), exceeding the 50 mrem/hr (500 µSv/hr) limit
specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal increase in radiation dose at thesite boundary. This was reported to the authorities at 2235.
The radiation dose at the main gate was 58.4 mrem/hr (583.7 µSv/hr), exceeding the 50 mrem/hr (500 µSv/hr) limit
specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal increase in radiation dose at thesite boundary. This was reported to the authorities at 0700.
Nonessential personnel were temporarily evacuated to Fukushima Daini. Government offices were informed.
Steam was observed issuing from the upper part of the reactor building. Government offices were informed.
The radiation dose at the main gate was 80.7 mrem/hr (807 µSv/hr), exceeding the 50 mrem/hr (500 µSv/hr) limit
specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal emission of radioactive materials
from a fire or explosion. This was reported to the authorities at 0836.
The Prime Minister directed that residents living between 12.4 and 18.6 miles (20 and 30 km) of the station take shelter.
The radiation dose at the main gate was 53.2 mrem/hr (531.6 µSv/hr), exceeding the 50 mrem/hr (500 µSv/hr) limit
specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal increase in radiation dose at thesite boundary. This was reported to the authorities at 1622.
Workers noted that the suppression chamber vent valve had closed.
The suppression chamber vent valve was opened.
The radiation dose at the main gate was 454.8 mrem/hr (4548 µSv/hr), exceeding the 50 mrem/hr (500 µSv/hr) limit
specified in Article 15, paragraph 1 of the Nuclear Disaster Law, indicating an abnormal increase in radiation dose at thesite boundary. This was reported to the authorities at 2320.
The bypass valve for the suppression chamber air-operated vent valve was opened.

97