Report of ‘Seminars to Investigate theAccident at the Fukushima Daiichi Nuclear
Power Station’
What went wrong? What should be done from now on?

March 2013
(Translation: February 2015)

Nuclear Safety Division
Atomic Energy Society of Japan

Report of ‘Seminars to investigate the accident at the Fukushima Daiichi Nuclear
Power Station’
What went wrong? What should be done from now on?
Summary
A massive earthquake, later named ‘the 2011 Off the Pacific Coast of Tohoku
Earthquake’ of moment magnitude 9.0 occurred at 14:46 on March 11, 2011. With theseismic motions and the tsunami caused by the earthquake, the Fukushima Daiichi
Nuclear Power Station (Fukushima Daiichi) in Fukushima Prefecture, which had six
boiling water reactors, experienced a station blackout accident, i.e., loss of allalternative power supply accident. At Units 1, 2, and 3, which were operative at the timeof the earthquake, the cores melted, and a large amount of radioactive materials werereleased into the environment.
To investigate this accident, the Nuclear Safety Division (NSD) of the Atomic
Energy Society of Japan (AESJ) organized a series of seminars from February 17, 2012
to December 19, 2012. This report summarizes the lessons learned from the eightseminars as well as the results of the discussion by the NSD executive members
throughout and after the series of seminars.
In the first seminar, overall issues regarding the accident were discussed. The
second to fifth seminars dealt with some specific issues found in Fukushima Daiichi,
namely, accident progressions, safety design, severe accident management, spent fuelpool safety, and emergency response: these issues were explained and discussed by theparticipants, including NSD members and AESJ members.
In the sixth and seventh seminars, the incidents that occurred in otherplants—Units 5 and 6 in Fukushima Daiichi, 4 units in Fukushima Daini Nuclear Power
Station (Fukushima Daini), 3 units in Onagawa Nuclear Power Station (Onagawa) and 1
unit in Tokai No. 2 Power Station (Tokai No. 2) —were explained and discussed.
Although these units were also affected by the severe seismic motion and tsunami
waves, severe accidents were prevented. The participants tried to understand the past
activities that contributed to the prevention of a severe accident. In the seventh seminar,also discussed was how Fukushima Daiichi was supported from the outside during theaccident, to identify the logistics issues. All the seven seminars had around 100
participants.

After the seventh seminar, NSD executive members started summarizing theissues mentioned in these seminars; they are described in this report. In the eighth
seminar, which had about 150 participants, the contents to be covered in the report wereexplained. The NSD executive members took into consideration the many comments
received during the eighth seminar and the discussions on various issues in preparing
this final report.
This report is organized as follows:- Chapter 1 presents an introduction.
- Chapter 2 describes the outlines of the seminars and defines the key words---

used in the report.
Chapter 3 explains safety design and severe accident management.
Chapter 4 describes the earthquake and the tsunami, as well as the preventive
measures incorporated against tsunami strikes in the past design process.
Chapter 5 describes the events involved in the accident at each unit of
Fukushima Daiichi, including accident progression, the estimation ofradioactive materials released into the environment, the logistics, emergency
responses, and International Nuclear and Radiological Event Scale (INES)
evaluation,
Chapter 6 compares the Fukushima Daiichi accident with the incidents
reported from the units of the other affected nuclear power stations, namely,
Fukushima Daini, Onagawa, and Tokai No.2.

-

Chapter 7 discusses a variety of safety-related issues, including defense indepth concept; safety design and severe accident management; nuclear
regulation; emergency preparedness; cooperation and communication among
relevant organizations; updating of accident-related information especially ofwhat really happened; probabilistic risk assessment (PRA) issues; operating
experience feedback; and safety research.
Chapter 8 presents the summary and conclusions of the report.

-

Chapter 9 gives the final statements.

The AESJ-NSD expects this report to be useful to understand well the accident
itself and the issues raised by the accident.

Condition of the Translation
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Summary

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Contents

Summary

1

Contentsi1.

Introduction

1

2.

Outline of the Seminar Series and Glossary of Technical Terms

3

3.

4.

2.1

Outline of the Seminar Series

3

2.2

Wording9

Summary of the Safety Design of Fukushima Daiichi

13

3.1

Safety Design of the Facility

13

3.2

Seismic Design and Tsunami-Proof Design

22

Tsunami Triggered by the Tohoku-Pacific Ocean Earthquake and Histories to

Determine Postulated Tsunami

25

5.

30

Summary of Fukushima Daiichi Nuclear Accident

5.1

Damage by Earthquake and Tsunami

30

5.2

Fukushima Daiichi Unit 1

35

5.3

Fukushima Daiichi Unit 2

46

5.4

Fukushima Daiichi Unit 3

56

5.5

Spent Fuel Pool

63

5.6

Fukushima Daiichi Unit 5, 6

72

5.7

Evaluation of the Released Radioactive Material

74

5.7.1

Radioactive Material Released into Atmosphere

74

5.7.2

Radioactive Material Released into Sea

75

5.8

Physical Distribution of Resources during the Accident

78

5.8.1 On-site Environment

78

5.8.2 Various Resource Supports for Fukushima Daiichi

79

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5.9

5.9.1

Radiation around Fukushima Daiichi during the Accident

85

5.9.2

Emergency Response

87

5.9.3

Initial Protective Action regarding Contamination of Foodstuffs and Water

88

5.9.4

Implementation of Additional Protective Actions

91

5.9.5

Estimation of Residents’ Exposure Dose

93

5.9.6

Protective Actions in Long-term Contaminated Areas

93

5.10

6.

85

Off-site Emergency Response

INES

97

5.10.1

INES Introduction

97

5.10.2

INES Rating of Fukushima Accident

97

5.10.3Issues in the INES Rating and International Correspondence99

Summary of Other Nuclear Power Plants

6.1

Fukushima Daini Nuclear Power Station

104

104

6.1.1

Outline of the Earthquake and Associated Tsunami

104

6.1.2

Influence of the Earthquake and Tsunami

105

6.1.3

Response After the Tsunami

106

Onagawa Nuclear Power Station

113

6.2

6.2.1

Summary of Earthquake and Tsunami, and Subsequent Responses

113

6.2.2

Damage to the Facility

117

6.2.3

Successful Operation and Efficacy of Safety Functions of Onagawa

119

6.2.4

Conclusion

120

6.3

Tokai No. 2 Nuclear Power Station

121

6.3.1

Summary of Earthquake and Tsunami

121

6.3.2

Impact of Earthquake and Tsunami

121

6.3.3

Situation at the Facility After the Earthquake

123

6.3.4

Countermeasures against Postulated Tsunami

126

6.3.5

Conclusion

127

6.4

Comparison of Events in Nuclear Power Plants

128

6.4.1

Viewpoint for Comparative Analysis

128

6.4.2

Seismic Motion Observed in Each Power Station and its Effects

129

6.4.3

Tsunami Wave Observed in Each Power Station and its Effects

132

6.4.4

Site Conditions after the Earthquake and Tsunami and the Recovery

Operation

138

6.4.5

142

Conclusion of the Comparison between Facilities

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Issues Identified by the Accident

7.1

Summary of issues According to DiD Concept

7.1.1
7.1.2

7.2

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Relationship between Safety Measures and the DiD
Particularly prevalent problems in the Fukushima Daiichi accident

Issues on Safety Design

7.2.1

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Event Analysis of the Fukushima Daiichi Accident, and the Lessons Related to

Safety Design

154

7.2.2

Summary of Issues and Measures, Concerning Safety Design on the Concept

DiD

157

7.3

Issues in Safety Regulations

164

7.3.1

Reconsideration of Principles in the Light of the Accident

164

7.3.2

Importance of Continuous Improvement within the Regulation

164

7.3.3

Issues Involved in the Design Basis related to Initiators such as Natural

Phenomena
7.3.4

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Establishment of Regulatory Requirements against Accident Management

Measures

168

7.3.5

Regulations using a Scientific Rationality are still Necessary

169

7.3.6

Issues Relating to the Independency of Regulations versus Collaboration

between Industries, Academia, and the Government

7.4

Issues on Nuclear Emergency Response

171

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7.4.1 Issues on the Implementation of Emergency Protective Actions

176

7.4.2 Issues on Crisis Management and Operation of Emergency Response

188

7.4.3 Conclusions

191

7.5

Issues on Cooperation and Communication during the Accident

193

7.5.1

Outline of the Problem

193

7.5.2

Organizational Roles when Severe Accidents Occur

193

7.5.3

Issues on Information Sharing during the Fukushima Daiichi Accident

196

7.6

Issues on the Updating of Information Relating to the Accident

202

7.7

Issues on the Utilization of PRA, Operating Experience, and Safety Research 205

7.7.1

Relationship between PRA, Operating Experience, and Safety Research

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7.7.2

PRA

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7.7.3

Operating Experience Analysis

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7.7.4

Safety Research

210

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Reference
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Conclusions

218

9.

Final Statements

223

Appendix 1: List of Abbreviations

224

Appendix 2: Contributors’ List

226

List of Draft Writers

227

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1.

Introduction
A massive earthquake of moment magnitude 9.0 occurred at 14:46 on March 11,
2011, in the region to the east of the Pacific Coast of Tohoku, Japan. The earthquake
was later named ‘The 2011 off the Pacific Coast of Tohoku Earthquake’. All the nuclear
power plants under operation at Fukushima Daiichi Nuclear Power Station (Fukushima
Daiichi) of Tokyo Electric Power Company (TEPCO) were automatically scrammed bythe ‘large seismic acceleration’ signals issued by seismic triggers. At approximately
15:30, a huge tsunami struck Fukushima Daiichi. The highest tsunami wave elevation
was far beyond the level considered in the design of the nuclear power station. Units 1
to 5 of the six nuclear power plants in Fukushima Daiichi experienced station blackout
accidents or accidents involving the loss of all alternating current (AC) power. All thereactors also lost the ultimate heat sink. Units 1 and 2 even lost their direct current (DC)power. The reactor cores of Units 1, 2, and 3 melted, releasing a large amount ofradioactive materials into the environment.
To investigate this accident, the Nuclear Safety Division (NSD) of the Atomic
Energy Society of Japan (AESJ) organized a series of ‘Seminars to investigate theaccident at the Fukushima Daiichi Nuclear Power Station’; eight seminars wereconducted from February 17 to December 19, 2012.
This report summarizes the presentations and the discussions conducted inthese Seminars. Several investigation reports of the Fukushima Daiichi accidents,
including the first and the second reports of the Japanese government and the reports bysome accident investigation committees, have been published. The AESJ is also
carrying out a variety of activities. This report presents an original analysis andevaluation by the AESJ-NSD members who are experts on nuclear safety. Some of thembelong to the organizations involved in the accidents directly or indirectly.
We would like to begin this report by quoting Mr. Kazuo Sato, former
Chairman of the Nuclear Safety Commission (NSC); in his lecture entitled “The Logicof Nuclear Safety” delivered at a Lecture Meeting organized by the NSD on February
17, 2010 in Tokyo, he said that the responsibility of operators and regulators is like apair of wheel. Neither can avoid their respective responsibilities regardless of theactions of the other.
The Fukushima Daiichi accident showed that neither the operator nor the
regulator had fulfilled their responsibility. Most of the NSD members, in particular, theNSD executive members, were deeply involved in safety-related activities in the past in
1their organizations, and hence, they are directly or indirectly related to the accident tocertain extents. In their capacity as responsible experts, they analysed the accident tofind out “what went wrong?” and “what should be done from now on?”The first draft of each section of the report was written by a suitable NSD
executive member or a lecturer at the Seminar. The drafts were finalized throughrepeated discussions by NSD executive members.
Accident progression is best known to the persons who were on the field.
Similarly, organizational problems may not be recognized well from outside of theorganization. Therefore, the name of the person in charge of drafting is shown at the endof each section. However, NSD executive members bear the responsibility for thecontents of the whole report.
Since we aim the report to be a compact, reader-friendly report, some of thedetailed discussions are not included in it. Readers can access more details on various
issues at the NSD webpage —http://www.aesj.or.jp/~safety/. NSD has uploaded thematerials (in Japanese) used for the presentations in the Seminars and the minutes of the
Seminars.
Nuclear technology is an integrated technology. Sometimes accidents,
including the Fukushima Daiichi one, occur at the interface between several technicalfields. In the seminars, the NSD aimed to achieve the following:・ Describe the accident progression as accurately as possible, by using the information
obtained at the time of writing the report
・ Identify what exactly went wrong by considering the reflections of the persons
involved in the accident,
・ Prescribe future steps to establish safer nuclear technology on the basis of thelessons learned from the accident
・ Enhance the mutual understanding among personnel from various technical fields aswell as among industry, academia, and the regulator.
The NSD expects this report to be useful to people related to nuclear safety.
The NSD also hopes that this report will contribute to realizing a better understanding ofaccident progression by other people, including the general public, journalists, and evenpeople who are against nuclear power. The NSD welcomes comments and criticisms on
our present and future works.
(Drafter: Kiyoharu Abe)
2

2.
2.1

Outline of the Seminar Series and Glossary of Technical Terms
Outline of the Seminar Series
NSD is the 17th of the 18 divisions of AESJ, and was established in August
2008. This division aims to contribute to the safe utilization of nuclear energy and torealizing good public understanding of nuclear safety by enhancing the interactions andcommunication between nuclear safety researchers, by supporting research activities,
and by disseminating information on nuclear safety.
The NSD is a gathering of individuals and experts whose occupations areclosely related to the accident. The NSD hence could not start its activity immediately
after the Fukushima Daiichi accident. Though the members of NSD felt the
responsibility for NSD, they were very busy in their own organizations with handlingthe accident and other related works. The NSD started a series of seminars on theaccident at about one year after the accident, recognizing its responsibility to provide aforum for sharing information and discussions among the members of AESJ.
The series comprised eight seminars, each of which had a specific theme fordiscussion. The themes are listed below. The titles and the names of the lecturers at eachseminar are listed in Table 2-1.
1st Seminar:
2nd Seminar:
3rd Seminar:
4th Seminar:th

5 Seminar:
6th Seminar:
7th Seminar:
8th Seminar:

Overall issues regarding the Fukushima Daiichi accident
Accident in Unit 1, and issues on safety design
Accidents in Unit 2 and 3, and issues on safety design
Accident in Unit 4, and issues on spent fuel pools; anintermediate summary of the seminar series
Emergency responses during the Fukushima Daiichi accident
and related issues
The situation in Onagawa and Tokai No.2
Accident at Fukushima Daini and Fukushima Daiichi Units 5
and 6
Final Summary

In the 1st Seminar, an overview of the accident progression at the Fukushima
Daiichi was presented. Five important topics were selected and discussed after thelecture: 1) safety design; 2) severe accident management; 3) off-site emergency
preparedness; 4) the use of probabilistic risk assessment, operating experience, andresearch findings; and 5) other issues, including that on plant life. The participants
expressed their appreciation of the seminar and requested the NSD to continue this3

valuable activity.
In the 2nd to 4th Seminars, the participants discussed safety design and severe
accident management related to the progression of the accidents in Units 1 to 4. In the
2nd Seminar, safety design was discussed by considering the basic principles of safety
design and the progression of the accident in Unit 1. In the 3rd Seminar, the design of the
instrumentation system and its credibility in the event of severe accidents werediscussed on the basis of the progression of the accidents in Units 2 and 3. In the 4th
Seminar, safety measures for spent fuel pool were discussed on the basis of the situationin Unit 4. The discussions carried out in the 1st to 3rd Seminar were summarized in the
4th Seminar.
In the 5th Seminar, off-site emergency responses and the lessons learned fromthe responses were discussed. The 6th and 7th Seminars dwelt upon Onagawa Nuclear
Power Station (Onagawa) of Tohoku Electric Power Co., Inc., Tokai No.2 Power Station(Tokai No. 2) of The Japan Atomic Power Company, Fukushima Daini Nuclear Power
Station (Fukushima Daini), and Fukushima Daiichi Nuclear Power Units 5 and 6.
Although these power plants also experienced severe seismic motion and were affectedby the tsunami, a severe accident was prevented. The activities at these sites, including
the preparedness against a tsunami before the earthquake, the damage inflicted by theseismic motion and the tsunami waves, and support activities, such as procurement ofneeded equipment, from the regions outside the site, were the topics of the presentations
and discussions held in the 6th and 7th Seminars. The participants tried to understand thefactors that contributed to the prevention of a severe accident.
Topics for the discussions were carefully selected by the NSD executive
members to avoid any bias resulting from the personal interests of the members. For
example, when talking about severe accidents, some experts may focus onphenomenology and simulation codes. However, in the event of a severe accident, weshould pay attention not only to such technological issues, but also to other important
subjects such as how to provide external support to the affected site during the accident,
how to assess the situation of the plant using limited information, and how to share
information with the cooperating organizations. The NSD executive members made aneffort to cover a wide range of subjects in the seminar series.
Eight seminars were conducted. Although this is a large contribution from theviewpoint of volunteer activity, it is not enough for an in-depth examination of various
issues related to the accident. Some issues were merely pointed out without offering any
solution.
All seminars followed almost the same pattern: The Chairperson was Naoto
4

Sekimura, Vice-President of the NSD. Individuals and experts, often NSD executive
members, involved in activities related to the accident delivered lectures first. Thereafter,
the issues to be discussed in the seminar were pointed by Kiyoharu Abe, the president of
the NSD. The discussion chaired by Sekimura was open for all the participants. The
seminars were managed by the NSD members from boiling water reactor (BWR) utility
group, pressurized water reactor (PWR) utility group, and the Japan Atomic Energy
Agency (JAEA).
(Drafter: Ken Muramatsu, Kiyoharu Abe)

Table 2-1 The programmes and speakers at the eight seminars (1/4)
No. (Date)

Theme, Title of Lectures

Lecturer (The Role in NDS, Affiliation)

No. 1

Report of “Seminars to investigate the accident at the Fukushima-Daiichi Nuclear Power(Feb. 17)

Station”(Discussing wide topics regarding to secure safety)
Naoto Sekimura (Vice-President, Todai※1)

1． Objective of the Seminar

2． Issues on Safety Design- Lessons and countermeasures learned from the Fukushima
Kumiaki Moriya (Hitachi-GE※2)

Accident
3． Issues on Severe Accident Countermeasures
Countermeasures in PWR Shigeru Urata(EC, KEPCO※3)

Countermeasures in BWR Tsutomu Matsui(EC, Chubu EPC※4)

4． Issues on Nuclear Emergenc] Response

Toshimitsu Homma (EC, JAEA)

5． Issues on PSA, Operational Experiences, and Safety Research
Discussion related to Safety Research

Toyoshi Fuketa (EC, JAEA)

6． Other Issues
Issues on Plant Life

Hiroshi Miyano (Hosei University)

No. 2

Accident in Unit 1, and Issues on Safety Design(May 8)

1． Objective of the Seminar

Naoto Sekimura (Vice-President, Todai)

2． Briefing of the Accident
Accident in Fukushima Daiichi Unit 1

Koichi Miyata (EC, TEPCO)

Safety Design in Fukushima Daiichi Unit 1, and Lessons Learned from the
Accident

Kumiaki Moriya (Hitachi-GE)

3． Safety Measures and Stress Tests
4． Discussion

Shigeru Urata (EC, KEPCO)

Chaired by Naoto Sekimura (Vice-President, Todai)

Summary of the Issues to be Discussed

5

Kiyoharu Abe (President, JNES※5)

Table 2-1 The programmes and speakers at the eight seminars (2/4)
No. (Date)

Theme, Title of Lectures

Lecturer (The Role in NDS, Affiliation)

No. 3

Accident in Unit 2 & 3, and Issues on Safety Design(June 26)

1． Objective of the Seminar

Naoto Sekimura (Vice-President, Todai)

2． Report of the Second Seminar

Ken Muramatsu (EC, Tokyo City University)

3． Briefing of the Accident and Raising the Problems
Accident in Fukushima Daiichi Unit 2 & Unit 3

Koichi Miyata (EC, TEPCO)

Issues on Instrumentation and Control Systems
4． Discussion

Koichiro Isoda (Toshiba)

Chaired by Naoto Sekimura (Vice-President, Todai)

Summary of the Issues to be Discussed
5． Introduction of Following Seminars

Kiyoharu Abe (President, JNES)
Takashi Nitta (Vice-President, JAEC※6)

No. 4

Accident in Unit 4, Issues on Spent Fuel Pool, and Mid-Term Summary(August 10)

1． Objective of the Seminar

Naoto Sekimura (Vice-President, Todai)

2． Report of the Third Seminar

Tsutomu Matsui (EC, Chubu EPC)

3． Accident in Fukushima Daiichi Unit 4 – The Situation in the Spent Fuel Pool
Koichi Miyata (EC, TEPCO)
4． Discussion

Chaired by Naoto Sekimura (Vice Chair, Todai)

Events in SFP and relating Issues to be Solved.

Kiyoharu Abe (Chair, JNES)

5． Safety Measures reflecting Lessons Learned from the Fukushima Daiichi Accident
Objective of Mid-Term Summary

Kiyoharu Abe (Chair, JNES)

Safety Measures reflecting Lessons Learned from the Fukushima Daiichi Accidents
Kumiaki Moriya (Hitachi-GE)
Discussion

Chaired by Naoto Sekimura (Vice-President, Todai)
Panelists: Takanori Tanaka (IEA※7), and Norio Watanabe (JAEA)

6． Introduction of Following Seminars

Takashi Nitta (Vice-President, JAEC)

No. 5

Issues on Emergency Response during the Fukushima Daiichi Accident(August 20)

1． Objective of the Seminar

Naoto Sekimura (Vice-President, Todai)
Narumitsu Umezawa (EC, MHI ※8)

2． Report of the Fourth Seminar

3． Issues on Emergency Response during the Fukushima Daiichi Accident
TEPCO’s Monitoring Activities during the Accident

Hiroaki Saneshige (TEPCO)

Role of Local Headquarter on Initial Activities

Minoru Saito (JNES)

Issues on Emergency Preparedness and Response

Toshimitsu Homma (EC, JAEA)

Issues on Emergency Response
4． Discussion

Kiyoharu Abe (President, JNES)

Chaired by Naoto Sekimura (Vice-President, Todai)

5． Introduction of Following Seminars

6

Takashi Nitta (Vice-President, JAEC)

Table 2-1 The programmes and speakers at the eight seminars (3/4)
No.

(Date)

Theme, Title of Lectures

Lecturer (The Role in NDS, Affiliation)

No. 6

Onagawa Nuclear Power Station and Tokai No. 2 Nuclear Power Station(October 22)

1． Objective of the Seminar

Naoto Sekimura (Vice-President, Todai)

2． Report of the Fifth Seminar

Toshimitsu Homma (EC, JAEA)

3． Situation at Onatawa Nuclear Power Station after the Tohoku-Pacific Ocean
Tatsuyuki Sakuraba (Tohoku EPC※9)

Earthquake

4． Situation at Tokai No. 2 Power Station after the Tohoku-Pacific Ocean Earthquake
Noriaki Tomura (JAEC)
5． Discussion

Chaired by Naoto Sekimura (Vice-President, Todai)

Summary of the Issues to be Discussed
Panelists:

Kiyoharu Abe (President, JNES)
Tatsuyuki Sakuraba, Noriaki Tomura

6． Introduction of Following Seminars

Takashi Nitta (Vice-President, JAEC)

No. 7

Fukushima Daini Nuclear Power Station and Fukushima Daiichi Unit 5~6(November 2)

1． Objective of the Seminar

Naoto Sekimura (Vice-President, Todai)

2． Report of the Sixth Seminar

Tsutomu Matsui (EC, Chubu EPC)

3． Situation at Fukushima Daini Nuclear Power Station
Kazuyuki Nagawasa (TEPCO)
4． The Accident at Fukushima Daiichi Unit 5 and 6

Koichi Miyata (EC, TEPCO)

5． Physical Distribution of Resources at the Fukushima Daiichi Accident, and thePoints for the Improvement of the Logistics
6． Discussion

Atsushi Kawamoto (TEPCO)

Chaired by Naoto Sekimura (Vice-President, Todai)

Summary of the Issues to be Discussed
7． Introduction of Following Seminars

7

Kiyoharu Abe (President, JNES)
Takashi Nitta (Vice-President, JAEC)

Table 2-1 The programmes and speakers at the eight seminars (4/4)
No.

Theme, Title of Lectures(Date)

Lecturer (The Role in NDS, Affiliation)

No. 8

Report of the Summary of the Seminar(December

1． Introduction

19)

2． Safety Design at Fukushima Daiichi Nuclear Power Station

Naoto Sekimura (Vice-President, Todai)

Kumiaki Moriya (Hitachi-GE)
3． Briefing of the Tohoku-Pacific Ocean Earthquake and the Fukushima Daiichi
Accident

Koichi Miyata (EC, TEPCO)

4． Briefing of Emergency Responses and Issues on Them
Mitsutoshi Homma (EC, JAEA)
5． INES Rating

Kiyoharu Abe (President, JNES)

6． Briefing of Incidents at the Other Nuclear Power Stations
Tsutomu Matsui (EC, Chubu EPC)
7． Issues Found at the Accident (1)Kumiaki Moriya (Hitachi-GE)

8． Issues Found at the Accident (2) and the Summary of the Seminar
Kiyoharu Abe (President, JNES)
9． Discussion

Chaired by Naoto Sekimura (Vice-President, Todai)

10． Introduction of Following Events

Takashi Nitta (Vice-President, JAEC)※1 Todai: The University of Tokyo;※2 Hitachi-GE: Hitachi-GE Nuclear Energy Ltd.
※3 KEPCO: The Kansai Electric Power Co., Inc.
※4 Chubu EPC: Chubu Electric Power Co., Inc.
※5 JNES: Japan Nuclear Energy Safety Organization※6 JAEC: The Japan Atomic Power Company※7 IAE: The Institute of Applied Energy※8 MHI: Mitsubishi Heavy Industry Ltd.
※9 Tohoku EPC: Tohoku Electric Power Co., Inc.

8

2.2

Wording
The Fukushima Daiichi Accident resulted in core melting and the release of alarge amount of radioactivity into the environment from Units 1 to 3. The accident wascaused by the loss of important safety functions because of seismic motion and tsunami
wave, which exceeded the levels considered in the design process. In the technicalanalysis of the accident, we should examine the following points:
・ The relationship between an initiator of accidents and the events that occur within
the facilities; in other words, how did the so-called “External Events” such as
seismic motion and tsunami wave initiate events such as transient and loss of
coolant accident (LOCA), loss of safety functions, and core melt
・ The response to each initiator, which exceeds the postulated levels in the design
process・ The response to events that occur within the facilities and exceed the postulated
conditions in the design process
Even prior to the Fukushima Daiichi accident, Japan’s approach for nuclear safety
covered some measures to cope with external events. However, most approaches are
based on the “Internal event” assumption—i.e. equipment failures occur randomly.
The Fukushima Daiichi accident suggests that we should pay more attention toexternal events; external events should be considered more directly and explicitly fromthe perspective of applying the “defense in depth” (DiD) concept to the concrete safetymeasures.
However, external events have been defined in different ways in the field of
nuclear safety. In addition, definition of the terms “Postulated Events” and “Beyond
Postulated Events” are not common among the personnel involved in nuclear safety.
Hence, we have defined some important technical terms at the beginning of the
discussion in this report:(1) Initiator, External Initiator, and Internal Initiator
Initiator: This is an event that may cause equipment failure in the facility.
Damage or failure of equipment results from two types of causes. Some causes
are inherent in the systems, structures, and components (SSCs); for example, ageing.
The other types of causes are external factors other than SSCs, such as seismic motion,
tsunami waves, and turbine missiles.
There are two approaches to discussing SSC failures. One is to consider thecausal relationship where the SSC failures are caused by some initiators. The other isbased on the assumption that SSCs fail randomly, and the frequencies of the failures9

are estimated using statistical data. The probabilistic risk assessment (PRA)
methodology is a typical example of these two approaches; the methodology for aso-called external event PRA considers the relationship between the external impacts
on SSCs and the SSC failures, while the methodology for a so-called internal event
PRA just assumes random failures. Initiators discussed in this report with emphasis on
the impact of external factors on SSCs because these initiators can cause simultaneous
failure of many SSCs.
The safety measures for various initiators should take into account thecharacteristics of the initiators. For example, seismic motion impacts all SSCs
simultaneously, so that all the safety-related SSCs are required to withstand seismic
loads according to their safety importance. On the other hand, the safety measuresagainst tsunami waves are quite different from those for other initiators—sufficiently
high elevation of the site ground, high elevation of seawall, and watertight doors forthe buildings.
External initiator: This is an initiator that occurs outside the facility.
There are two types of external initiators—”Natural Events” and “Man-madeEvents”. Natural events include seismic motion, tsunami waves, strong winds, storm
surges, volcanic ash, and pyroclastic flow. Man-made events include accidental andintentional aircraft crashes, occupation of the main control room, and cyber-attacks.
Note that it is almost impossible to control the “Hazard” (defined later in the report)of external initiators.
Internal initiator: This is an initiator that occurs within the facility.
Some internal initiators, such as ageing, are inherent in the SSCs themselves.
Other internal initiators are external impacts on the SSCs. Internal fire, internal
flooding, and turbine missile strikes are typical examples of such initiators. This
report focuses mainly on the latter. Unlike the external initiators, it is possible tocontrol the hazard of internal initiators.
We have avoided using popular terms such as “Internal Event” and “External
Event” because their definitions are not always commonly used among the relevant
people.
For example, in PRA, internal fire and internal flooding are sometimes
categorized as external events because they are impacts of external factors on SSCs, andthese events are analysed by the “External Event PRA Method”, while they are10sometimes categorized as internal events because they occur within the facility or
because their hazards can be controlled.
Internal event PRAs are usually performed by assuming random failure ofSSCs, and the methodologies is different from fire and flooding PRAs. Although theterms “Internal Event PRA” and “External Event PRA” are widely used in the field of
PRA research, it may be more appropriate to use the terms “Random Event PRA” and“Initiator-Specific PRA”, respectively.
(2) Hazard, Postulation, and Design Basis
Hazard: This refers to the magnitude (or strength or height) of each initiator.
Probabilistic hazard: This refers to hazard described as a relationship between its
magnitude and frequency, or the description of frequency as a function of each levelof magnitude, or the description of magnitude as a function of each level of exceededfrequency. In general, the larger the magnitude, the smaller is the frequency.
Design basis hazard (DBH): These are hazards against which the soundness of theequipment is assured by the safety design. Translation of “Design Basis” into
Japanese is not common for all the initiators but the concepts are almost common.
The DBH of seismic motion is “Design Basis Seismic Motion”. The DBH of tsunami
wave was called “Postulated Tsunami Height” until the Fukushima Daiichi accident,
because the concept of “Design Basis” from the perspective of a tsunami was unclear
until the accident. The DBH of terrorist attack is “design basis threat,” or “DBT”.
Even if a certain initiator exceeds its DBH, it does not indicate the occurrenceof a so-called level 4 event (beyond design basis accident) of the DiD concept. For
example, when an earthquake occurs, the seismic motion at the facility may exceed
the design basis seismic motion. However, because of the large safety margins in theseismic design, if the reactor does not experience any serious damage except for atransient resulting from a reactor trip, which is a normal result of the actuation of theseismic triggers, the event is ranked as level 2 according to the DiD concept.
Design basis seismic motion: This is the magnitude of seismic motion that thefacility can withstand owing to the seismic design.
Postulated tsunami wave: This is almost the same as the design basis tsunami wave.
This is the height of tsunami wave that is considered in a tsunami proof design. As
11mentioned above, the word design basis tsunami was not used until the Fukushima
Daiichi accident. (Nuclear Regulation Authority (NRA) now uses the word “Design
Basis Tsunami Height” in the new regulatory requirements.) This report uses the oldwording, “Postulated Tsunami Wave” because it investigates the problems before theaccident.
(3) Design Basis Event and Design Basis Accident
Design basis event (DBE): An event that is postulated to review the adequacy of thesafety design and the safety assessment of a facility and its SSCs.
The former NSC prescribed the “Regulatory Guide for Reviewing Safety
Assessment of Light Water Reactors (Safety Assessment Guide)”, to examine theoverall safety of a facility where two levels of design basis abnormal
conditions—anticipated operational transients and accidents—were postulated. The
NSC also prescribe the “Regulatory Guide for Reviewing Safety Design of Light
Water Reactors (Safety Design Guide)” to examine the adequacy of each SSC design.
According to the Safety Design Guide, each SSC should fulfil its safety function
under the postulated situations—DBH events, such as design basis seismic motion, as
well as DBE events, such as postulated transients or accidents. If any equipment, suchas a filtered vent system, is expected to be robust to severe accident conditions
according to its safety design, the specific severe accident condition is also defined as
the DBE for that equipment.
Design Basis Accident (DBA): A DBA is a type of DBEs. DBAs are defined in the
NSC Safety Assessment Guide.
(4) Acronyms
Acronyms are listed at the end of this report.
(Drafter: Kiyoharu Abe)

12

3.

Summary of the Safety Design of Fukushima Daiichi
The safety design of the Fukushima Daiichi is explained in this chapter. Our
objective is to introduce the background information to help understand the progressionof the accident and to discuss the lessons learned from the accident. Section 3.1
explains the safety system and equipment related to the accident. Section 3.2 explains
the seismic design and tsunami-proof design.
3.1

Safety Design of the Facility
Nuclear power plants are designed to ensure safety; for this purpose, they are
equipped with multiple and diverse systems and equipment based on the defense indepth philosophy. The design condition must be defined during the actual design
process by postulating representative scenarios that cover various accidents and byconsidering sufficient margins against these scenarios.
The term “beyond the postulation”, which has been frequently used after theaccident, should be reworded as “beyond design postulation”. It is evident that design
postulation has limitations in the actual design. The minimum safety function must be
secured by risk management, even if the hazard exceeds the design postulation.
Therefore, we examined the residual risks that are not covered by the postulated
scenario. And accident management (AM) measures, such as preparing additional water
injection system using systems that are not originally equipped for safety purpose, hadbeen introduced.
As all units of Fukushima Daiichi are BWRs whose direct cycle basic
management strategy is to feed water into the reactor core through many diverse
injection systems.
(1) Safety Equipment of the Plants
Fukushima Daiichi has six BWR plants. Unit 1 is an early-type BWR called
BWR-3, which is described in Figure 3-1. Units 2–5 are BWR-4 plants, which is
described in Figure 3-2. The primary containment vessels (PCVs) of these five plants
are of MARK-I type, as shown in Figure 3-3. Unit 6 is the only BWR-5 in Fukushima
Daiichi, and its PCV is of MARK-II type, as shown in Figure 3-4. Specifications ofUnits 1–4 are summarized in Table 3-1. In the following subsections, the isolation
condenser (IC) and the other reactor injection equipment, the PCV and the related
system, the hydrogen countermeasure in the PCV are described, and instrumentation forreactor core water measurement is explained.
13

a. Isolation Condenser and the Other Reactor Injection Equipment
The safety equipment of BWR-3 is shown in Figure 3-1. The safety system forreactor isolation incident comprises two ICs. The emergency core cooling system(ECCS) comprises one turbine-driven high-pressure core injection system (HPCI) andtwo motor-driven core splay systems (CSs). Four safety relief valves (SRVs) areprovided for depressurizing.
An IC is a heat exchanger that removes the decay heat when a reactor isisolated in the transient; it performs this function by condensing residual steam andreturning the condensed water into the reactor core. Since it has no injection pump, thecooling capability of IC is 200% of the decay heat after a scram corresponding to 3% of
the rated thermal output. An automatic start and a large margin of cooling capability are
adopted to ensure for suppressing reactor pressure for preventing loss of coolant by SRV
actuation. Soon after automatic activation of two ICs, the reactor pressure starts
decreasing quickly because of the large coolant capacity of the ICs. Therefore, plant
operators use only one valve of an IC and keep the other closed; this is also necessary inthe case of the operating IC in order to ensure that the reactor pressure does not decrease
too quickly, because the 100% capacity of an IC is designed under a conservative
assumption of decay heat. To reduce the number of operational request to the plant
operator, later plants adopted the reactor core isolation cooling system (RCIC). It feeds
water from outside of reactor pressure vessel (RPV) through a steam-driven turbine.
Residual steam is released into a suppression pool through an SRV.
If a break is detected in an IC line, or if the system loses its controlling power,
four motor-driven valves on each IC line are automatically closed by the interlock.
If an IC fails, the reactor pressure increases because of decay heat. SRVs areautomatically opened and the residual steam is relieved into suppression pool to prevent
overpressure. It leads the reduction of reactor core water level. According to the signalof “Low Reactor Water Level”, the HPCI starts automatically injecting water fromoutside. Even if the HPCI failed to start operating, the automatic depressurization
system (ADS) starts automatically depressurizing the reactor according to the signal of“Low-Low Reactor Water Level”. The CSs start operating simultaneously with the ADS
to spray water into the core.
The safety equipment of BWR-4 like Units 2–5 is shown in Figure 3-2. A
turbine-driven RCIC is installed, instead of an IC, as a safety measure for reactor
isolation incidents, as mentioned above. The ECCS comprises a low-pressure core
injection system (LPCI) in addition to an HPCI and two CSs. Eight SRVs are provided,
as required for the output.
14

The cooling capacity of an RCIC is also designed under a conservative
assumption of decay heat. The water level in the reactor is raised by operating the RCIC.
To prevent mist/steam flowing into the main steam line, RCIC is automatically trippedby the signal “High Reactor Water Level”.
The diversity in injection systems was a basic strategy in the AM guideline
prepared by the former NSC around 2000. Injection into reactor core and PCV was
safeguarded by remotely controlled feeding systems using fire protection lines and acondensate storage tank (CST), as shown in Figure 3-5.
b. PCV and the Related System
A PCV forms a boundary to contain radioactive materials in the event of an
accident. All lines penetrating a PCV usually have two isolation valves; of these, one
isolation valve is set on the inside of the PCV, and the other is set on the outside. The
protection of the PCV boundary usually is assigned priority over the other safetyfunctions. Hence, a fail-safe design is adopted for the isolation valves. They close
automatically when any abnormal event occurs. In the case of a feeding system such as
an ECCS, however, the feeding function has the same or higher priority over theisolation function. In this case, check valves are used instead of isolation valves, which
are provided inside a PCV, on the feeding line.
c. Hydrogen Countermeasure in a PCV
An atmosphere inside PCV is replaced into nitrogen gas before a plant start
operation. According to previous experiments, if the oxygen concentration in a PCV isless than 5%, hydrogen undergoes combustion. This helps prevent hydrogen combustioneven if hydrogen is generated by the oxidation of the fuel cladding made of a zirconium
alloy under high temperature during an accident. Since hydrogen and oxygen are
generated slowly by water radiolysis, the PCVs are equipped with a Flammability
Control System (FCS) to recombine the generated hydrogen and oxygen. In the case of
the Fukushima Daiichi accident, it took at least 40–50 h for the concentration of
inflammable gases to reach the flammability limit. Hence, although no FCS was activeat Fukushima Daiichi, radiolysis cannot be considered as the cause of the hydrogen
explosion that occurred about 40–50 h after station blackout.
Hydrogen countermeasures in reactor building were not considered in plantdesign before the Fukushima Daiichi accident.
d. Reactor Core Water Measurement
15

The water level in a reactor core is measured by considering the difference inthe water head, as shown in Figure 3-6. The water head is measured at two points. Oneof the points is along a line in contact with the liquid phase in the RPV, whose waterhead shifts with the water level of the reactor core. The other point is as a stablereference point in a condenser in contact with the steam phase of RPV. This mechanism
enables us to observe the water level directly, and any changes can be quickly andaccurately determined. However, calibration of indicated water level is required when
the water density changes because of shifts in reactor pressure.
This mechanism is, however, not reliable under conditions of overheating. The
water level of the reference condenser cannot be maintained stable when the reactor
overheats and/or when the temperature in the PCV became higher than the saturation
temperature of the reactor. Under these conditions, the water in the reference condenser
starts evaporating. This leads to an overestimation of the water level. Under prolonged
overheating, the water in the line connected to the water phase of reactor also starts
evaporating. This also leads to errors in water level measurement, with lower levels
being measured with increasing evaporation. To counter the situation like this, AM
procedures wherein water measurements are unavailable are prepared.
e. Accident Management Countermeasure
AM countermeasures were installed in all the existing nuclear power plants inJapan around the year 2000 to reduce the risk of events beyond the design basis. The
basic strategy of the countermeasure involves the use of diverse water injection systems,
heat sink systems, and electrical power sources. An alternate water injection system anda PCV venting system are shown in Figure 3-5 as examples of AM countermeasures.
An AM injection system feeds water from a filtrate water tank and/or a CST.
The water sources are dispersed in a nuclear power station.
A PCV venting system protects the PCV boundary from overpressure failure.
However, AM countermeasures were installed on the condition that they do not affectthe safety systems in a DBA, so that the rapture disk, whose set point is higher than the
maximum allowable pressure of the PCV, was installed on its line to maintain theboundary function.
Sharing of power sources is another AM measure. Both AC and DC powersources are connected to a neighbouring unit by sharing lines.
f. Layout of Equipment
Systems and equipment that have safety function were located as dispersedly16and independently as possible. Pumps and other apparatus of the ECCS are separated
systematically and are installed in each divisional room based on the system separation.
A diesel generator for use in emergencies is installed in the basement and/or in the
ground floor because of its heavy weight.

(2) Emergency Operation Procedures
In addition to the equipment, the emergency operation procedure is very
important. In any BWR accidents, the emergency operation procedure involves
depressurizing the reactor, injecting water, and finally removing the decay heat after thescram. Operators try to ensure that these procedures are followed by using diverse andmultiplied equipment explained above.
The fundamental measure for BWR is to inject water into reactor core in any
accidents involving the loss of coolant. As described in Figure 3-7, the reactor isdepressurized, while the coolant water is supplied by high-pressure injection systems,
such as the IC, RCIC, and HPCI. Thereafter, low-pressure injection systems inject water
into the reactor. The safety equipment and AM equipment are operated automatically
and/or manually.
(Drafter: Kumiaki Moriya)

17

Fig. 3-1 The main safety systems of BWR-3

Fig. 3-2 The main safety systems of BWR-4

18

Fig. 3-3 MARK-I containment vessel

Fig. 3-4 MARK-II containment vessel
19Fig. 3-5 Examples of system response based on accident management(First: Alternative water injection system; Second: Pressure-strengthened
containment vent)20

Fig. 3-6 Principle of reactor water level indicator

Fig. 3-7 Basis response to accident

21

3.2

Seismic Design and Tsunami-Proof Design
Defense design for both internal and external initiators follows the common
philosophy based on initiation frequency. If the initiator has a small frequency of lessthan 10-7 events per year, it can be excluded from the safety design. If the frequency ishigher, protection design against this initiator is a must. A hazard with a frequency of
10-4 or 10-3 events per year is defined as a DBH, and protection with large safety margin
has to be considered. If the frequency of severe accidents resulting from by an initiation
event is sufficiently low, e.g. lower than the target for a safety goal, the protection
designs for such initiators are considered to be appropriate.
(1) Regulatory Guide for Reviewing Safety Design of Light Water Nuclear Power
Reactor Facilities
The regulatory requirements for external initiators were summarized on thebasis of the case study of application for reactor establishment permit of Fukushima
Daiichi Unit 3’s.
The safety design was reviewed by the Reactor Safety Examining Board of theJapan Atomic Energy Commission.
“Regulatory Guide for Reviewing Safety Design of Light Water Nuclear Power
Reactor Facilities” (Safety Design Guide) was published in 1970. This guide wasestablished by observing “General Design Criteria for Nuclear Power Plant
Construction Permits” (General Design Criteria) published by the United States Nuclear
Regulatory Commission (USNRC) in July 1967.
Safety Design Guide requires that the most severe natural load be consideredwhen designing for the natural condition. The guide also requires that a graded approachagainst seismic motion be adopted. Equipment should be separated by considering theeffect of the seismic hazard and classified based on the importance of seismic safety.
(2) Seismic Design
According to the application for reactor establishment permit of Fukushima
Daiichi, its seismic design was based on the following policy. A rigid structure isadopted. Important buildings, such as the reactor building, are supported directly by the
bedrock. Equipment is classified and designed according to its importance for seismic
safety.
The seismic importance classification is as follows:
Class As:The equipment included in class A and possessing a function
for emergency response; e.g. the PCV and control rods drive.
22

Class A:

Class B:The equipment whose failure has a potential to cause severe
accident and/or which prevents a hazard to the public; e.g. thereactor building and reactor pressure vessel.
The equipment related to high-level radioactive materials; e.g.

Class C:the radioactive waste treatment building and radioactive
waste treatment equipment.
The equipment not included classes As, A, or B.

(3) Tsunami-Proof Design
Tsunami-proof design was developed observing the General Design Criteria
established by the USNRC. The highest tide level had been already decided as a design
requirement in the basic design stage of Fukushima Daiichi. The application for reactor
establishment permit of Fukushima Daiichi stated that the highest tide level was set at+3.122 m from Onahama Point (O.P.), which is a reference water surface level for
Fukushima Daiichi. The value was defined on the basis of the tide level of Chile
Tsunami in 1960.
In the aftermath of the Hokkaido south-west offshore earthquake of 1993, across-ministry review of the tsunami countermeasure was triggered in Japan. The
tsunami evaluation of nuclear power stations was reviewed by a board of the Japan
Society of Civil Engineers (JSCE), and a new evaluation method was established. The
new method assumes the highest level of tsunami waves that may be triggered by greatinter-plate earthquake. These developments were adopted in the revision of theRegulatory Guide for Reviewing Seismic Design for Nuclear Power Reactor Facilities
in 2006. Tsunamis are mentioned as events associated with earthquakes in the revised
guide, and the guide requires that important safety function of a nuclear facility not be
severely damaged by tsunami.
(4) Summary of Seismic Design and Tsunami-Proof Design
The protection requirements for natural disasters, including seismic motions
and tsunami, are described in the application for the establishment permit. The nuclear
facility has been classified according to its importance, and important equipment was
designed with large safety margins. As new knowledge emerged, the relevant equipment
has been re-evaluated by nuclear utilities consulting with regulatory body, and has beenstrengthened to increase the seismic safety margin. However, both safety margin forseismic and conventional tsunami countermeasures, which are setting high enough site
elevation and breakwater, was based on the design postulation of the hazards. A review
23

of the previous countermeasures in the light of Fukushima Daiichi accident revealed
that no concrete countermeasures against hazards beyond the postulation had been
adopted.
(Drafter: Kumiaki Moriya)

24

4.
Tsunami Triggered by the Tohoku-Pacific Ocean Earthquake andHistories to Determine Postulated Tsunami(1) Tsunami triggered by the Tohoku-Pacific Ocean Earthquake
The 2011 off the Pacific Coast of Tohoku Earthquake (hereafter, theTohoku-Pacific Ocean Earthquake) occurred at 14:46 on March 11, 2011. Its seismic
origin was off the Sanriku coast, at latitude 38° N and longitude 142.9° E. Figure 4-1
shows the source area of the earthquake.
This earthquake was an inter-plate earthquake that occurred along the Japanese
Trench, and it had several seismic sources, including the ones ‘off the Miyagi
Prefecture’, ‘south part of the off the Sanriku coast’, ‘off the Fukushima prefecture’, and‘off the Ibaraki prefecture’. The observed magnitude of 9.0 made it the fourth largest
earthquake in the world.
The highest seismic intensity of 7 was measured in Kurihara City in Miyagi
Prefecture. The intensities recorded in the municipalities where nuclear power plants arelocated are as follows: intensity 6 lower in Onagawa, Miyagi; intensity 6 upper in
Okuma and Tomioka, Fukushima; and intensity 6 lower in Tokai, Ibaraki.
Subsequently, at around 15:30, the huge tsunami struck Onagawa, Fukushima
Daiichi, Fukushima Daini, and Tokai No. 2.
The maximum tsunami heights at these nuclear power plants were O.P. +13.0
m at Onagawa (O.P.: the tide level at Onagawa point) and H.P. +5.5 m at Tokai No. 2(H.P.: the tide level at Hitachi point). At Fukushima Daiichi and Fukushima Daini, thetsunami heights were not recorded because the tide gauges were damaged; however,numerical analysis indicated that the tsunami heights were about O.P. +13 m at
Fukushima Daiichi and O.P. +9 m at Fukushima Daini (here O.P. stands for the sea levelat Onahama Point that is the standard sea level for both Fukushima Daiichi and
Fukushima Daini). Immediately after the tsunami stuck, an announcement was made to
the effect that a 15 m tsunami struck the site, but this height was the run-up height.
Before 2011, the Postulated tsunami height at the nuclear power stations had
been evaluated on the basis of “Tsunami Assessment Method for Nuclear Power Plants
in Japan” (2002), issued by the JSCE. According to this method, the region from the
Pacific coast to Japan Trench is divided into several wave sources, and the height oftsunami is evaluated for each wave source.
The tsunami triggered by the Tohoku-Pacific Ocean Earthquake exceeded thepostulates of the JSCE’s tsunami evaluation method in several aspects: (i) more than25

one wave sources were activated together; (ii) tsunami waves overlapped because of thetime difference in plate rupturing; and (iii) the slippage of the plate sliding also
exceeded the postulation.
Figure 4-2 shows the wave source model considered in the JSCE’s tsunami
evaluation method. Figure 4-3 illustrates the wave source model of the Tohoku-Pacific
Ocean Earthquake generated by inversion analysis to reproduce the recorded tsunami.
(2) Histories to Determine Tsunami Postulation at NPPs
When the license applications of Fukushima Daiichi and Fukushima Dainiwere submitted, the Japanese regulatory body did not have any clear standards on thetsunami protection of nuclear power plants. Hence, these plants were designed based on
the largest documented tsunami that had occurred along the Fukushima coast—thetsunami caused by a 1960 Chilean earthquake. In 2002, JSCE developed a new method
for evaluating the postulated tsunami height; by this method, the postulated tsunami
height was calculated to be 5.7 m for Fukushima Daiichi and 5.2 m for Fukushima
Daini. Upon this evaluation, measures were implemented to protect the nuclear power
stations from a large tsunami; seawater pumps were raised, and water-tight doors were
installed in buildings whose elevation was lower than the postulated tsunami height. In
2007, Fukushima and Ibaraki Prefectures published an update on the postulated tsunami
height; however, the updated values were lower than the previous ones evaluated by
JSCE’s method, and hence, no additional action was taken. In 2009, the postulated
tsunami height increased from 5.7 m to 6.1 m after the re-evaluation of the seabedtopography and tide level data; hence, the elevation of the seawater pumps wasincreased once again.
In addition to the evaluation of the postulated tsunami height, continuous
efforts have been made to reflect newly obtained information on tsunami into nuclear
power station design. One example of such efforts was the way in which a statement
issued by the Headquarters for Earthquake Research Promotion (HERP) in 2002 washandled: although small friction and strain accumulation were not expected to cause alarge earthquake along the inter-plate off Fukushima which triggered the earthquake thistime, HERP stated that this type of earthquake could occur even in this region. However,
the wave source model for this statement was not verified at the time and the HERP’s
did not suppose the earthquake caused by more than one seismic source. Considering
HERP’s statement, TEPCO started working on the development of the probabilistic risk
assessment (PRA) method for tsunami. They simultaneously conducted a trial
evaluation by postulating a wave source model with characteristics of Meiji-Sanriku-oki
26Earthquake and asked the JSCE to review the appropriateness of the wave source
models.
Around the same time, TEPCO received the research literature on the Jogan
earthquake and tsunami published by the National Institute of Advanced Industrial
Science and Technology. TEPCO then conducted a trial calculation and investigation oftsunami deposits in order to confirm the effect of the Jogan tsunami. The calculation
result assuming the worst scenario was O.P. +8~9 m at the front of intake point of the
Fukushima Daiichi even though the coupling of more than one seismic source was notconsidered. As a result of the investigations, tsunami deposits from Jogan tsunami wereconfirmed at Minami-Soma City in the northern part of Fukushima Prefecture; notsunami deposits were found in the southern part between Tomioka and Iwaki.
The content of this section is a summary of what was reported in the Seminar 4-1,
4-2, 4-3).
(Drafter: Koichi Miyata)

Reference
4-1
N. Tomura, “Situation at Tokai No. 2 Nuclear Power Station after the
Tohoku-Pacific Ocean Earthquake”, Lecture slides for the 6th Seminar
4-2
K. Nagasawa, “Situation at Fukushima Daini Nuclear Power Station”, Lectureslides for the 7th Seminar
4-3

Tohoku Electric Power Company, “Situation at Onagawa Nuclear Power
Station after the Tohoku-Pacific Ocean Earthquake “, Lecture slides for the 6th
Seminar27

Fig. 4-1 The Focal Area of the Tohoku-Pacific Ocean Earthquake

Fig. 4-2 JSCE’s wave source and Jogan tsunami’s wave source28

Fig. 4-3 The reproduction analysis model of the tsunami triggered by
Tohoku-Pacific Ocean Earthquake by TEPCO

29

5.
5.1

Summary of Fukushima Daiichi Nuclear Accident
Damage by Earthquake and Tsunami
Figure 5.1-1 shows that the earthquake magnitude experienced at Okuma
Town and Futaba Town where Fukushima Daiichi is located was 6 upper. This
earthquake was caused by several seismic sources. Until this earthquake, theHeadquarters for Earthquake Research (HERP) had not even considered the possibilityof such a large earthquake caused by more than one seismic source in this region.
According to the seismic record of the seismic indicators installed in the undergroundfloor of the reactor buildings of each Fukushima Daiichi unit, the shaking was slightly
above or equal to that corresponding to the design basis earthquake ground motion Ss as
shown in Table 5.1-1.
As a result of the seismic motion, a transmission line tower collapsed, andcircuit breakers were damaged. External power was lost in the entire power station
given in Figure 5.1-2, and the reactors at operation were automatically stopped (scram).
With the loss of external power supply, all 12 emergency diesel generators (EDG)
except the one in Unit 4 that was under inspection automatically started, and power
supply was ensured. At this stage, based on the plant parameters such as reactor waterlevel, reactor pressure, and primary containment vessel (PCV) temperature, it was
deduced that there was no damage in reactor coolant pressure boundary. In addition, a
few low seismic class equipment incurred damage as shown in Figure5.1-3, but thisdamage had no effect on reactor safety. The detailed future investigation is expected toassess the impact of earthquake on major safety equipment.
A large tsunami stuck about an hour after the earthquake. The tsunami height
was estimated by TEPCO as about 13 m near the tide station where the design basis
tsunami height was defined at design stage as shown in Table 5.1-2. Although TEPCO
reported immediately after the tsunami that the height was above 15 m, this height wasthe ‘run-up height’. The JSCE had created a tsunami evaluation model consideringvarious tsunami sources. However, this tsunami, which is the fourth largest tsunami
recorded worldwide, was caused by several wave sources and not just a single one asshown in Figure 5.1-4. The Japan Meteorological Agency issued a 3-m-high tsunami
warning initially and updated the information twice; a >10 m tsunami warning wasissued at 15:30.
The tsunami flooded the Fukushima Daiichi facilities to a large extent, andmost seawater systems were damaged. Water entered major buildings, and many power
panels were lost as seen in Table 5.1-3. All the AC power in Units 1–5 and DC power in30

Units 1, 2, and 4 were lost.
In Fukushima Daiichi, as a countermeasure against station blackout (SBO), theDC power supply required for the IC, the RCIC, and the HPCI and a high-voltage
power connection with an adjacent unit were originally equipped. Before the earthquake,a low-voltage power connection was established with the adjacent unit to prepare for the
event of loss of DC power as well. However, since the adjacent unit also lost its power,
the initial condition of the plants was extremely severe. In addition, large aftershocks,
the continued tsunami warning, and the tsunami debris in the site made it difficult toinitiate the required response to the accident.
Detailed information on the damage caused by the earthquake and tsunami at
Fukushima Daiichi is recorded in the reference [5-1], and the details of this seminar are
also given in references [5-2].
(Drafter: Koichi Miyata)Fig. 5.1-1 Distribution of Seismic Intensity of the Tohoku-Pacific Ocean
Earthquake

31

Table 5.1-1 Record of Observed Accelerations and Maximum Response
Accelerations against Design Basis Earthquake

Fig. 5.1-2 Status of External Power Supply System

32

Fig. 5.1-3 Piping around Moisture Separator Heater(Small Diameter piping diverged from No. 3 Drain Piping was raptured)

Table 5.1-2 Summary of Tsunami Information

33

Fig. 5.1-4 Wave Sources of the Tsunami

Table 5.1-3 Damage in Support Systems

34

5.2

Fukushima Daiichi Unit 1
The earthquake caused the loss of off-site power and an automatic reactor
scram. Two EDGs started automatically, and the main steam isolation valve closed.
Hence, the pressure in the RPV increased, the ICs were automatically initiated, andreactor water level was maintained.
Both the trains of the IC system operated well as designed, and the pressure
and temperature in the RPV decreased. However, in order to prevent the reactor cool
down rate from exceeding the technical specification rate, the operator manually started
and stopped the IC system thrice to control the cooling.
After earthquake, but before tsunami struck, although the off-site power was
lost, the necessary safety functions responded well as designed, and the reactor coreremained stable.
After the tsunami struck, in addition to the loss of the seawater cooling system,
the buildings that was located 10 m above sea level was flooded. The flooding wetted orsubmerged EDGs and power distribution systems, resulting in a gradual loss of all AC
and DC power. The loss of power caused a loss of normal control room lighting,indicators, and controls. All the instrumentation necessary for monitoring andcontrolling the emergency became unavailable.
According to TEPCO’s investigation, as a result of the loss of DC power, theisolation interlock of the ICs functioned, and the isolation valves closed. This fail-safe
design of isolation interlock caused the loss of IC cooling function.
Just after tsunami, it was unknown whether the IC valves were open or closed.
At about 18:00 on March 11, the operator confirmed that the closure indication light(green light) was on when the DC power was restored temporarily, and then the operator
opened the IC valves. At that time, Fukushima Daiichi’s ERC erroneously considered
the IC system functioning and that the water level in the RPV was maintained. However,the site superintendent anticipated that water injection would become necessary andordered water injection using fire engines.
In fact, contrary to the operator’s knowledge, the IC system had stopped
functioning owing to the impact of the tsunami. The RPV pressure increased, opening
the SRVs, and steam condensed in the suppression chamber (S/C). As a result, the waterlevel in the RPV decreased. Because of the loss of DC power, the plant parameters
could not be obtained after the tsunami. The water level indication came buck at 21:19
on 11 March by connecting portable batteries. Figure 5.2-1, 5-2-2, and 5-2–3 shows thewater level and pressure deduced by severe accident analysis code (MAAP) byconsidering the plant conditions and operator actions as per TEPCO’s investigations.
35

The analysis shows that the water level reached under the top of active fuel(TAF) after 18:00 on March 11, and core damage started before 19:00. The field RPVpressure indicator was confirmed as 7.0 MPa [abs] at around 20:00, which would be theresult of the actuation of the safety valve function of SRVs, and at this stage, the reactor
coolant pressure boundary seemed intact.
After 21:00, the measured water level was above TAF; however, there is a big
difference between the measurement and analysis results. Later investigations by
TEPCO showed that after the reactor core was exposed, the water in the condensing pot
evaporated, and water level indicated at the time of the event was erroneous. When theactual water level was decreasing continuously, the water in the reactor side piping
started evaporating, and the actual water level indication seemed to be responding tothis mechanism. By 21:00, the reactor core had already been exposed and core damage
had started. This caused the measured water level to be higher than the actual level.
At about 00:00 on March 12, the pressure in the dry well (DW) was 0.6 MPa[abs], which exceeded the design pressure of 0.531 MPa [abs]. For instance, in the
LOCA, the maximum pressure was 0.401 MPa [abs]; that is, the reactor core seemed to
be in an abnormal condition. Upon receipt of this information, the site superintendent
ordered the preparation of the containment vent. At 02:30, the DW pressure increased to
0.84 MPa [abs], while the pressure in RPV was almost the same—0.8 MPa [gage].
These pressure responses suggest that there was some leakage in the reactor coolant
pressure boundary until at the time. Core damage continued until this time, and thetemperature in the reactor increased.
According to TEPCO’s MAAP analysis under the assumption of dry tube and
SRV gasket failure as shown in Figure 5.2-4, the RPV pressure was equalized with the
PCV pressure, as shown in Figure 5.2-2. This analysis shows that the RPVs weredamaged by the effect of core melt debris at about 02:00 March 12.
As the DW pressure reached approximately 0.75 MPa [abs], radioactive
materials might have been released. In fact, in the early morning on March 12, the dose
rate measured by the monitoring car located near main gate increased; thus, it was clearthat radioactive materials were released from Unit 1.
At around 04:00 on March 12, water injection was initiated, as shown in
Figure 5.2-5, using a fire engine. At the beginning, fresh water from the fire protection
tank was used, but the supply was soon exhausted, and seawater was used. The injection
path used here (fire protection system (FP) → Make up water system (MUWC) → Core
spray system (CS)) had been installed in accordance with AM. Though diesel-driven
36

fire pump (DD-FP) could not be used because the fuel supply was exhausted, a fireengine was used to pump water into FP. The fire engine was placed in Fukushima
Daiichi as the lesson from Chuetsu-oki Earthquake in 2007, by which FP system at
Kashiwazaki-Kariwa got unavailable.
Although water injection could not prevent core damage, injected watersupposed to reach the core melt debris that had dropped onto the containment pedestal.
According to MAAP analysis, molten core concrete interaction (MCCI) was suppressed;
the melt debris corroded the concrete pedestal by approximately 70 cm, and it was
retained there as shown in Figure 5.2-6.
Along with water injection, operators began preparations for containment
venting. Because AC power and compressed air were lost, the only way for containment
venting was to open two valves manually, as shown in Figure 5.2-7.
The motor-driven valve on the second floor of reactor building was opened
manually. However, the operators needed to access the underground torus room to open
the other air-operated S/C vent valve. After 9:00 on March 12, the team entered torus
room. 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 open the valve. They could not reach to thevalve because the dose quickly exceeded the predetermined limits. One of the operators
received a dose exceeding the emergency dose limit of 100 mSv. In order to open theseair-driven valves remotely, compressed air and DC power were necessary to actuate thesolenoid valve. However, compressed air was not available. The operators attempted toopen the small valve from the control room by relying on the residual air pressure in the
instrumentation air system using a portable generator. After 10:00, the radiation level
increased at the main gate monitoring car, as shown in Figure 5.2-9. However, the PCV
pressure remained high, indicating that the venting was not fully effective. It was
unclear whether the rupture disk on the vent line was open or not.
To supply compressed air, a small air compressor was brought to reactor
building equipment bay and connected to the line. By 14:00, the compressor was
installed and was started. At 14:30, almost 24 hours after the event initiation, the PCV
pressure decreased and steam release was confirmed as shown in Figure 5.2-10. The
containment vent was considered successfully operated; however, the radiation level at
the main gate did not increase at the time.
At 15:36 on March 12, the reactor building of Unit 1 exploded as shown in
Figure 5.2-11, and radiation level at the main gate increased temporarily.
37

Various factors such as reactivity accident, steam explosion inside and outside
the reactor, and direct containment heating, were considered to explain the explosion of
the building. However, even after the explosion, positive pressure remained in the PCV,and the explosions inside the PCV were not suitable explanation of this event.
Hydrogen generated from the high-temperature interaction of zirconium andsteam was being released from the reactor into the PCV. Some of these gases found theirway into the reactor building; it is difficult to determine the path of leakage, thoughsome potential paths have been identified.
Since the explosion happened just an hour after containment venting, it isnecessary to figure out the relationship between containment venting and the leakage ofhydrogen gas from the PCV. The venting pipes were connected a stand-by gas treatment
system (SGTS). Although the entrance of the SGTS was isolated from the containment
vent line by an isolation valve, most of the valves along the SGTS line are kept open
when electric power is lost, as shown in Figure 5.2-12. A flow-control damper, which
was kept close on losing power, was installed on the reverse path from the venting
piping. This damper can limit the reverse flow. However, since the leak-tightness of thedamper is insufficient, it could only restrain large flows. Besides, if the vent gas flowed
along the reverse path through the filter train of the SGTS, most of the radioactive
particles should be held up; this was inconsistent with the fact that the reactor building
was widely and heavily contaminated. Therefore, although it was difficult to fully deny
the reverse flow from the SGTS, the SGTS probably was not the main path along whichhydrogen flowed into the reactor building.
Detailed information on the Fukushima Daiichi Unit 1 accident is recorded in
the reference [5-1], and the details of this seminar are also given in references [5-2, 5-3].
(Drafter: Koichi Miyata)38

Fig. 5.2-1 Reactor Water Level (Unit 1)

Fig. 5.2-2 Reactor Pressure (Unit 1)39

Fig. 5.2-3 PCV Pressure (Unit 1)

Fig. 5.2-4 The Leakage Path from the RPV
40

Fig. 5.2-5 Water Injection Line to RPV

Fig. 5.2-6 Status of Fuel Debris (Unit 1)41

Fig. 5.2-7 Containment Vent ValvesFig. 5.2-8 Access Route for S/C Vent Valve (Small Valve)42

Fig. 5.2-9 Monitoring Data and PCV Pressure on March 12, 2011

Fig. 5.2-10 Steam Discharge from Unit 1, 2 Stack at 15:00 on March 12, 201143

Fig. 5.2-11 Damage to the Reactor Building (Unit 1)

44

Fig. 5.2-12 Possibility of SGTS Back Flow (Unit 1)

45

5.3

Fukushima Daiichi Unit 2
The earthquake caused a loss of off-site power and an automatic reactor scram.
Several events, including the automatic starting of two EDGs and the automatic closingof the main steam isolation valve, occurred as expected because of the loss of off-site
power. SRVs were initiated to maintain pressure in the RPV. When the water level
reaches a low level (L2), the RCIC starts automatically; however, in reality, theoperators manually initiated the RCIC before the water level reached L2. The RCIC
stopped upon receipt of the high water level (L8) trip, and the operator repeated themanual initiation of the RCIC, thus maintaining the water level under control before thetsunami struck.
The subsequent tsunamis not only affected the seawater cooling system, butalso flooded major buildings located 10 m above the sea level. Most of the power panels
in Unit 2 were lost. Control room lighting, including emergency light, failed completely,leaving control room in darkness, and most instrumentation necessary for monitoring
and controlling the emergency became unavailable.
Just before the tsunami struck, operator manually initiated the RCIC, but theoperating conditions thereafter are not clear. Because the water level in the reactor core
was also unknown, preparations for water injection were started at around 21:00 on
March 11 as per the AM procedures, and the alternative line had been prepared by the
midnight. At around 1:00 on March 12, operators attempted to verify the condition of
the RCIC. The RCIC room was dark, and the inundated water level in the room camenearly to the top of the worker’s boots. Hence, they turned back without getting closeenough to check system operation. At about 2:00 on March 12, workers made anotherattempt to verify RCIC operation. They found that the reactor pressure was 5.6 MPa,and the RCIC pump discharge pressure was 6.0 MPa; thus, they concluded that the
RCIC was operating.
At 22:00 on March 11, workers used batteries to restore reactor water levelindication. The indicated water level was above the TAF. The water level remained+3.4~3.9 m. Figure 5.3-1 shows the measured water level and the level calculated byusing MAAP code. The reactor water level remained high until the daytime on March
14. In addition, although the relief function of SRV did not work, the reactor corepressure was relatively low at around 6 MPa [gauge], as shown in Figure 5.3-2.
Although the RCIC is not expected to function when DC power is lost, in reality, theplant parameters remained stable. Subsequent analysis explains that thus stability was a46

result of the special operation of the RCIC as described below.
First, water level data from the indicator in the fuel zone were obtained. The
data had to be adjusted for high-pressure (6 MPa [gage]) conditions because the fuelzone indicators were adjusted to atmospheric pressure. After adjustment, the proofed
water level was equal to the height of the water level indicator condenser nozzle. If theactual water level were higher than this nozzle, considering the principle of the waterlevel indicator based on pressure difference, the measured water level would stay at
nozzle height level. Hence, the actual water level was supposed to be above the nozzle.
At this water level, a two-phase flow goes to the main steam line, and this flow drives
the RCIC turbine. The RCIC was not working efficiently since the injection flow was
lower than the designed value, and the two-phase flow and injection flow were balanced.
If the two-phase flow, which has more enthalpy per unit volume than does the steam,enters the main steam line, the RPV pressure is lower (6 MPa [gage]) than the openingpressure of SRVs.
The measured pressure in the PCV increased slower than expected upon theloss of heat removal function as seen from Figure 5.3-3. It was possible that some heatremoval mechanism was working. Water was injected towards Units 1–3 with damaged
reactor cores, but there was no evidence that the water level in the PCV increased. The
contaminated cooling water flowed into the torus room and turbine building of theseunits as well. Though Unit 4 did not experience the process by which flooding occurredin Unit 1-3, a large quantity of water was found in the torus room of Unit-4. Thissuggests that the tsunami inundated the turbine building, and the water flowed into thetorus room. If this was the situation in Unit 2, the torus could be cooled from outside.
MAAP analysis was conducted under the above assumption, and the calculated PCV
pressure agreed well with the measured data.
The measured pressure in the PCV was lower than the rupture disk designpressure of 0.531 MPa [abs], and venting preparation was carried on. Until 11:00 on
March 13, two valves on the vent line had remained open. Later, after the explosion of
Unit 3 on 14 March, the air-operated valves closed due to the failure of solenoid on the
air supply valve.
Preparing for the loss of the RCIC function, water injection alignment wascompleted at around 12:00 on March 13. Besides, around an hour later, batteries wereconnected with the SRV control panel in control room making depressurization andwater injection possible. However, debris from Unit 3 explosion damaged the fire47

engine and hoses that had been set up to inject seawater into reactor. As in the case of
Unit 1, the injection line (FP → MUWC →LPCI) was set up in accordance with AM
procedures, and the fire engine had been placed as the lesson of Chuetsu-oki
Earthquake.
During daytime on March 14, the reactor water level began to decrease,probably because of RCIC failure. Workers attempted to rebuild the water injection line
and containment vent path. Because of delays in opening the S/C vent valve, TEPCO
changed the priority from venting containment to depressurizing the reactor through the
SRV and water injection using the fire engine. Efforts to depressurize the reactor
continued until around 18:00, when the RPV pressure was successfully decreased. Highradiation levels on site, caused by radioactive debris from explosions of Units 1 and 3,
precluded the continuous staffing for fire engines; hence, fire engine operation waschecked intermittently. At 19:20, a worker touring the area discovered that the fireengine had run out of fuel. After replenishing fuel, the engine was restarted, and asecond fire engine injected water at around 20:00. Thereafter, the RPV pressure
increased a couple of times, and the workers attempted several times to open the SRVs.
Unit 2 was in an unstable condition.
According to MAAP analysis conducted by TEPCO, at around 17:00 on March
14, the reactor water level decreased to below TAF, and core damage started at around
19:20. Later, although the measured reactor water level increased temporarily,
considering the experience gained from the Unit 1 incident, the water level indicator
became unreliable after core was damaged. In the analysis, water injection flow wasassumed to be less than measured at the fire engine, as shown in Figure 5.3-4. The
duration from the time taken to attain a low water level to the time when water injection
was carried out was shorter than that in the case of Unit 1; hence, it was analysed that
the RPV in this case was not as damaged as in Unit 1. However, the investigation basedon plant parameters suggested that the RPV was damaged and that part of the melt
debris dropped down onto the pedestal region, as shown in Figure 5.3-5.
TEPCO personnel suspected that the solenoid on the air supply valve had failed,preventing the large air-operated S/C vent valve from opening. At 21:00 on March 14,
the small S/C air-operated vent valve was slightly opened. The indicated PCV pressure
remained slightly lower than the design pressure, and venting seemed unsuccessful. At
21:20, an operator opened a second SRV to encourage depressurization. The radiationlevel near the main gate increased temporarily as shown in Figure 5.3-6. However, it48

was not clear whether the rupture disk on the vent line opened or not. Thereafter, theDW pressure increased sharply, and the indicated pressure exceeded 700 kPa [abs] at
07:20 on March 15. Thus, it was deduced that the hydrogen produced by core damage
resulted in high pressure in DW. At past 0:00 on March 15, a small DW vent valve was
opened for a while, and several minutes later, it closed again. There was no pressurechange in the DW, and the monitoring car did not detect any change in the radiationlevel. Hence, the steam in the DW did not seem to be released outside. In addition, after
22:00 on March 14, gamma radiation was detected by the containment atmospheremonitoring system (CAMS). Gamma radiation increased synchronously with theprogress of core damage.
At past 06:00 on March 15, a loud noise was heard, and vibrations were sensed.
Meanwhile, S/C pressure indication failed (0 kPa [abs]). At this moment, there weredoubts regarding the intactness of S/C of Unit 2, and the nonessential personnel
evacuated to Fukushima Daini.
The indicated S/C pressure had slowly reduced since the midnight of March 14,
but DW pressure had not been accompanied by it. Later investigation by TEPCO
revealed that the S/C pressure reading of 0.0 MPa [abs] was an indication of a failed
instrument, and not an indication of vacuum. Further, a review of the seismic
instrumentation revealed that the shock resulted from the explosion of Unit 4 as shownin Figure 5.3-7; this was determined by considering the linearity between the arrival
time of the P wave and S wave and the distance from the object unit. Later, TEPCO
confirmed that the photographs of the Unit 2 torus room captured by a robot showed nosign of explosion, as given in Figure 5.3-8.
In other words, the simultaneous loud noise and failed S/C pressure indication
led to the misjudgement that the S/C had been damaged. This issue will also be
discussed in Section 7.6.
At 07:20 on March 15, the indicated DW pressure remained stable atapproximately 730 kPa [abs]. At 11:25, it indicated a lower value of 155 kPa [abs]. At
10:00 on March 15, white smoke rose near Unit 2, as observed from the photos taken by
a live camera installed in Fukushima Daiichi. Besides, the dose rate near the main gate
also increased dramatically, as shown in Figure 5.3-10, as a result of the release of agreat amount of radioactive material. TEPCO’s investigation later showed that the highradiation dose rate was observed around the shield plug at a reactor building operating
floor as shown in Figure 5.3-11. On the basis of this fact and by considering the49potential leakage path studied in a severe accident research program, TEPCO estimatedthat the seal of the PCV head flange was the main path of the leakage of radioactive
material.
In Unit 2, the reactor core was damaged, but the situation differed from that ofUnits 1 and 3 in that no hydrogen explosion occurred in the reactor building. It isthought that because the gas containing hydrogen was released from the blow-out panel
opening (4 × 6 m) caused by the explosion in Unit 1, there was no prolonged
accumulation of a high concentration of hydrogen in the reactor building of Unit 2.
The detailed information regarding Fukushima Daiichi Unit 2 accident isrecorded in the reference [5-1], and the detailed of this seminar is also given inreferences [5-3, 5-4].
(Drafter: Koichi Miyata)Fig. 5.3-1 Reactor water level (Unit 2)50

Fig. 5.3-2 Reactor pressure (Unit 2)

Fig. 5.3-3 PCV pressure (Unit 2)51

Fig. 5.3-4 Water Injection Flow Rate and analysis Condition (Unit 2)

Fig. 5.3-5 Status of Fuel Debris (Unit 2)52

Fig. 5.3-6 Monitoring Data and PCV pressure on March 14, 2011

Fig. 5.3-7 Estimation of explosion position based on records of seismograph53

Fig. 5.3-8 Status of S/C

Fig. 5.3-9 Steam Discharge from the Building54

Fig. 5.3-10 Monitoring Data and PCV Pressure on March 15, 2011

Fig. 5.3-11 Dose Distribution on the Fifth Floor of the Reactor Building
55

5.4

Fukushima Daiichi Unit 3
The earthquake caused the loss of off-site power and an automatic reactor
scram. Several events occurred; for example, two EDGs started automatically, and themain steam isolation valve closed. These were expected, considering the loss of off-site
AC power. SRVs were initiated to maintain the pressure in the RPV. When the waterlevel is low (L2), the RCIC starts automatically; in reality, however, the operators
manually initiated the RCIC after the scram before the water level decreased to L2.
The subsequent tsunamis affected the seawater cooling system and flooded
major buildings located 10 m above the sea level. Most of power panels in Unit 3 werelost, but some DC power remained available; this marks the difference between thesituations in Unit 3 and those in Units 1 and 2. Therefore, facilities depending on DC
power, such as the monitoring of reactor water level, RCIC, and HPCI still worked. Two
EDGs tripped, and then station went into a SBO condition. The RCIC remained
functional for water injection to control the water level. In order to save the DC power,part of the RCIC discharge water was flowed back to the CST, which was the water
source for the RCIC.
At 11:36 on March 12, the RCIC shutdown unexpectedly and could not berestarted. At this time, the reactor water level became low, and the HPCI automaticallystarted on receiving an L2 reactor water level signal. The operation of the HPCI was
similar to that of the RCIC. With the continued operation of the HPCI, the reactor waterlevel remained stable, as shown in Figure 5.4-1.
Figure 5.4-2 shows the indicated reactor water level and analysed result by
MAAP code. To operate continuously, the HPCI system was continuously drawing asignificant amount of steam off the reactor, and the reactor pressure decreased
considerably, as shown in Figure 5.4-3. At that time, the DW pressure was higher than
the analysed result, as shown in Figure 5.3-4. After 12:00 on March 12, the DD-FP wasoperated to spray the S/C, and then, the DW pressure became similar to the analysed
result; the reason for this is not clear.
Before the HPCI stopped operating, depressurization using SRV and alternative
water injection path using DD-FP, which was prepared as AM, was on standby. As theindicated reactor pressure became extremely low at 02:00 on March 13, operators
doubted that HPCI system was damaged and feared that reactor steam would be
released into atmosphere. At 02:42, the HPCI system was manually shutdown. At 02:45,
operators attempted to open SRV valves, but failed. With the HPCI isolated and with no
SRVs available, the reactor pressure quickly increased. They next planned to attempt56water injection by DD-FP, but the reactor pressure was higher than the dischargepressure of the fire pump, preventing water injection. Operators tried to open SRVs byusing car batteries. At around 9:00, SRVs opened to depressurize the reactor. Then,water injection using the fire engine and DD-FP began.
MAAP analysis assumed that water injection was stopped when operator
stopped HPCI operation. The analysed result revealed that past 9:00, the reactor waterlevel decreased to below TAF, and core damage started at around 10:40. However, theindicated water level was lower than the analysed water level, indicating that the waterlevel decreased to below TAF earlier than the calculated time and that core damage hadalready started before the estimated time. Thereafter, although the indicated reactor
water level increased temporarily, the water level indicator became unreliable after coredamage, as seen in the case of Unit 1. In the analysis water injection flow was assumed
to be less than that measured at the fire engine, as shown in Figure 5.4-5.
When the HPCI system was still working to inject water, DW pressure waslower than the design pressure to open a rupture disk. The vent line was lined up until
8:41 on March 13 so that venting would be initiated. if the rupture disk opened. At 9:20,
DW pressure decreased, indicating successful venting, as shown in Figure 5.4-6. Just a
moment after the first venting, the radiation level near the main gate increased
temporarily to around 300 Sv/h, but after the second and subsequent ventings, there
was no indication of increase in radiation. Although it is possible that depending on thewind direction, the dose level at the location of the monitoring car did not increase; if alarge amount of fission products was released through the ventilation stack, direct
radiation would have been observed. Since this was not the case, the fission products
released after the second and subsequent ventings were estimated to be limited. Before
the first venting, at around 8:00, the radiation dose was observed to increase atmonitoring point MP-4. At that time, DW pressure in Unit 3 was under the maximum
limit. Hence, the source of these fission products was probably Unit 1, whose core wasalready severely damaged.
Several measures were suggested in order to prevent hydrogen explosion; theseincluded opening of the blow-out panel for reactor building ventilation, and drilling
holes on the roof of the reactor building. When workers were preparing to bore a holeby using water jets, which reduces the probability of triggering an explosion, and before
the arrival of water jet equipment, the reactor building exploded at 11:01 on March 14,
as shown in Figure 5.4-7. This explosion was more severe than the one in Unit 1. The
smoke generated in this explosion was black, while that generated in the explosion of57

Unit 1 was white, which is the signature colour of hydrogen explosion. The top floor
walls of the reactor building in Unit 1 were steel construction, while those in Unit 3
were made of reinforced concrete. The black smoke is supposed a result of mixing of
crushed concrete. The reason behind the bigger explosion is presumed that more
explosive materials were released and Unit 3 has a stronger building structure than Unit
1. The cause of explosion, as in the case of Unit 1, was the leakage and accumulation of
the hydrogen produced by reactor core damage. Between core damage and explosion,
workers had attempted containment venting four times. The first venting was carried out
one day before explosion, and the last venting was conducted 4 hours before theexplosion. In the SGTS of Unit 3, dampers were installed on the reverse path from theventing pipe as shown in Figure 5.4-8, prohibiting a reverse current. These dampers
might not have been well sealed, though they were able to prevent a large flow rate.
TEPCO investigated the SGTS filter train on 22 December 2011 and. the detected dose
rate was very low (only a few mSv/h). If there had been a large reverse flow ofhydrogen and radioactive material, most of the radioactive particles would have been
trapped in the filter; however, the reactor building was broadly and heavily
contaminated. Hence, although it is difficult to fully deny that there was no reverse flowin the SGTS, it was not the main path through which hydrogen entered the reactor
building.
The detailed information on the Fukushima Daiichi Unit 3 accident is recordedin the reference [5-1], and the details of this seminar are also given in references [5-3,
5-4].
(Drafter: Koichi Miyata)58

Fig. 5.4-1 Operating Condition of HPCI

Fig. 5.4-2 Reactor Water Level (Unit 3)59

Fig. 5.4-3 Reactor Pressure (Unit 3)

Fig. 5.4-4 PCV Pressure (Unit 3)60

Fig. 5.4-5 Water Injection Flow Rate and Analysis Condition (Unit 3)

Fig. 5.4-6 Monitoring Data and PCV Pressure on March 13, 201161

Fig. 5.4-7 Damage of the Reactor Building (Unit 3)

Fig. 5.4-8 Possibility of SGTS Back Flow (Unit 3)
62

5.5

Spent Fuel Pool
The entire core had been offloaded into the spent fuel pool (SFP) to support
ongoing outage work in Unit 4. Hence, Unit 4 SFP had the greatest heat load among the
SFPs of Units 1-6 and a common SFP, as shown in Table 5.5-1.
Owing to the earthquake, off-site power was lost, SFP cooling system tripped,
and EDG (B) was automatically initiated. EDG (A) was under maintenance.
When the tsunami struck, all AC and DC power was lost, and SFP cooling
function was lost. It was estimated that fuel would be uncovered around late Marchbecause of decay heat. During the accident, the condition of Units 1–3 was so seriousthat addressing their core cooling had precedence. At around 4:00 on March 14, watertemperature in SFP was measured as approximately 84 °C.
At 6:12 on March 15, explosion occurred at the fifth floor of the reactor
building with a loud noise and vibrations, as shown in Figure 5.5-1. Since all the fuels
were in the SFP in Unit 4 and the water temperature was less than boiling temperature
as measured on March 14, hydrogen was not expected to be produced by themetal-water reaction. It was difficult to figure out the reason for the explosion at thattime. Moreover, the effect of the explosion on the pool and fuel were unknown, and thewater level in the SFP where there was no water level indicator was unknown. Station
personnel took numerous actions to ensure that the spent fuel was sufficiently cooled. If
there had been no water in SFP, the dose rate around Unit 4 caused by sky shine would
have been considerably high. However, the dose rate was moderate, enabling personnel
to work near Unit 4; hence, it was deemed that the fuel was not exposed.
On March 16, helicopter-borne observations revealed that water still remainedin the SFP. Fresh water was added using a water cannon truck on March 20, andseawater was introduced into the pool using a concrete pump March 22 onwards. Freshwater was once again added, instead of seawater, into the pool March 30 onwards. On
July 30, an alternative cooling system was introduced.
Various causes of the explosion were suggested: (a) hydrogen produced by theover-heated spent fuel; (b) gasified oil; (c) brought-in combustible gas; and (d)hydrogen produced by water radiolysis. None of these causes was verified owing to thefollowing reasons: (a) the SFP contained sufficient amount of water; (b) the temperaturein the building was not high enough to gasify oil; (c) the amount of combustible gas in
the gas cylinder was not sufficient for producing a large-scale explosion; (d) sufficienthydrogen was not expected to have been produced. The most likely explanation was thatthe hydrogen produced during core damage contributed to this explosion, and the63potential path was from Unit 3 via a pathway to the Unit 4 refuelling floor, flowing
through a shared pipe from PCV to the stack as shown in Figure 5.5-2. The SGTS
system in Unit 4 did not have dampers to suppress reverse flow. To confirm theassumption that hydrogen generated in Unit 3 flowed into Unit 4 via the SGTS, on
August 25, TEPCO performed a radiation survey on the SGTS filters. The survey result
revealed higher radiation levels at the locations closer to Unit 3 as shown in Figure5.5-3. These survey results support the conclusion that there was a backflow from Unit
3 to Unit 4. Thus, the hydrogen explosion in the Unit 4 reactor building is supposed tohave been caused by hydrogen flowing in from Unit 3.
According to the on-site investigation on the fourth and fifth storeys of thereactor building, the fifth storey’s floor had deformed upward, while the fourth storey’s
floor had deformed downward. The ventilation duct located on the fourth storey wascrushed into rubble and scattered as shown in Figure 5.5-4. Therefore, it was concluded
that the explosion occurred in the ventilation duct on the fourth storey and propagated to
the whole building through the stair hall.
Since there was no sign of water leakage on the second floor under the SFP andsince no damage was found in the SPF support structures, the structural integrity of SFP
was maintained. Post-accident simulations revealed that the reactor building, including
the SFP, could even endure an additional seismic event of the same magnitude as the
Tohoku-Pacific Ocean Earthquake. In order to ensure additional seismic margin, theSFP was reinforced from underneath as shown in Figure 5.5-5. An inclining survey ofthe reactor building was continuously performed. As shown in Figure 5.5-6, thedistance between the pool surface and floor, and the inclination of the building wall
were measured. The investigation showed that there was no significant inclination.
The photograph of the interior of the SFP did not show any fuel rack damage,and the fresh fuels taken from the pool were found to be intact. Moreover, theradiochemical assessment of the SFP revealed that iodine and caesium indeed existed,
but their concentrations were lower than those in Units 1–3 by two orders as shown in
Table 5.5-2. Although the concentrations were higher than those before the accident, theabsolute value was small, and SFP did not incur any severe damage. The detectedfission products most likely came from the damaged reactor cores of Units 1–3.
Since the confirmation of the water level in the SFP on March 16, much effort
had been made to inject water into the SFP by using water cannon and concrete pumptrucks. Water level measurements were made using concrete pump trucks since April 12.
64

The evaluation results of SFP water level and temperature at Unit 4 are shown in Figure5.5-7 along with the measured values. The evaluation assumed evaporation and thesloshing resulting from the earthquake and explosion. The amount of water that
evaporated had surpassed injection until around on 20 April, and the water level reached+1.5 m of the top of fuel rack. The pool gate was thought to be closed by water injection
on April 22, and water level recovered to full capacity by April 27 when the water levelin the skimmer surge tank increased largely.
The SFPs in Units 1–3 were assessed in the same manner as in the case of Unit
4. In comparison with the SFP in Unit 4, the decay heat was less, and the pooltemperature was as high as 70 °C. Although the water level decreased because ofevaporation, there was sufficient water in the SFPs of Units 1–3. In addition, although
the reactor buildings exploded in Units 1 and 3, the water-keeping function seemed to
be intact. The decay heat in the SFPs in Units 5 and 6 and the common pool was lessthan that in Unit 4, and they were maintained by introducing cooling facilities, though
the temperature once increased up to 70 °C. Table 5.5.-3 presents the radionuclide
analysis results of the pool water samples from Units 1–3. Although more than sevenmonths had passed after shutdown, short half-life radionuclides were detected, and theradionuclide composition was similar to those of the accumulated water in theunderground rooms of turbine buildings. Hence, the SFP may have been contaminatedby the damaged reactors.
With regard the dry cask storage facility, a large amount of seawater, sand, anddebris flowed into the building; the floor got wet, and the louvers and doors weredamaged. However, the dry storage casks were air-cooled and well-sealed as theradiation dose level was maintained at the background level.
The detailed information on SFPs is documented in the reference [5-1], and thedetails of this seminar are also given in references [5-5].
(Drafter: Koichi Miyata)65

Table 5.5-1 Spent Fuel Assemblies
Unit

Spent fuel assembly(New fuel)Decay heat (MW)(March 11)

Amount of waterin the pool (m3)Unit 1292 (100)

0.18

990

Unit 2587 (28)

0.62

1390

Unit 3514 (52)

0.54

1390

Unit 41331 (204)

2.26

1390

Unit 5946 (48)

1.01

1390

Unit 6876 (64)

0.87

About 1450

Common Pool

6375 (－)

1.13

About 4000

Cask Storage
Building

408 (－)－－

Fig. 5.5-1 Damage State of the Reactor Building (Unit 4)66

Fig. 5.5-2 Flow Path of Hydrogen Gas from Unit 3 to Unit 4

Fig. 5.5-3 Possibility of SGTS Back Flow (Unit 4)67

Fig. 5.5-4 Damage State of R/B due to Explosion (Unit 4)68Fig. 5.5-5 Anti-earthquake Reinforcement

Fig. 5.5-6 Tilt of the Reactor Building (Unit 4)69

Table 5.5-2 Sampling Results (Unit 4)

Fig. 5.5-7 Calculation Results of Water Temperature and Water Level of SFP (Unit4)

70Table 5.5-3 Results of samplings (Unit 3)

Fig. 5.5-8 Dry Storage Cask

71

5.6

Fukushima Daiichi Unit 5, 6
Units 5 and 6 were under refuelling outage and cold shutdown. Unit 5 wasunder pressure leakage testing of RPV, and the reactor pressure was maintained at about
7 MPa [gage]. The RPV head was on in Unit 6. The seismic motion at the both units
was same or slightly higher comparing with the design basis. The tsunami wave at Unit
5-6 significantly exceeded their design basis and the latest evaluation at that time. As
shown in Figure 5.6-1, however, both units suffered relatively less damage since their
ground levels were higher (O.P. +13 m) than that of Units 1-4 (O.P. +10 m).
The earthquake caused the loss of off-site power at Unit 5; hence, two EDGs
were automatically activated. After the control rod drive (CRD) pump tripped owing toloss of power, the reactor pressure temporarily dropped to about 5 MPa [gage].
Subsequently, owing to the impact of the tsunami, high voltage power panels andemergency low voltage power panels were fully lost, leading to SBO. Plant parameters
were available because some regular low voltage power panels and DC power supply
were intact. Since the reactor core had just been reloaded, the decay heat was relatively
low, and the reactor pressure increased slowly. At night on March 11, workers began theinvestigation and recovery of power equipment as shown in Figure 5.6-2. From March
12, the RPV was depressurized through its top vent. Until March 14, the nitrogen supply
system, which was kept out of service for the pressure leakage testing of RPV, wasrecovered, and the SRVs were manually opened intermittently to depressurize the RPV.
After power supply to the MUWC system was made available, water injection wasstarted and reactor water level was maintained on March 14. A portable underwater
pump was connected and was activated on March 19; the Residual Heat Removal
System (RHR) (C) system was then restored to cool the SPF. On March 20, reactor
cooling started via the RHR (C) system, and the reactor attained cold shutdown. Since
then, the SFP and reactor core were cooled alternately by the RHR system.
The earthquake caused loss of off-site power at Unit 6, and then three EDGs
were automatically activated. The subsequent tsunami made part of the high voltage
power panels unavailable; however, DC power was maintained, and one air-cooleddiesel generator (6B) was also intact, preventing SBO in Unit 6. Cooling water injection
via MUWC started on March 13. Reactor pressure increased gradually owing to decay
heat. Starting March 14, the SRV was initiated to depressurize the reactor, and thereactor pressure was maintained low enough for MUWC to enable water injection into
the reactor. A portable underwater pump was activated on March 19; the RHR (B)72system was restored to cool the SPF. On March 20, reactor cooling was started using the
RHR (B) system, and the reactor attained cold shutdown. Since then, the SFP andreactor core were cooled alternately by the RHR system.
The detailed information on Fukushima Daiichi Units 5 and 6 is documented in
the reference [5-1], and the details of this seminar are also given in references [5-6].
(Drafter: Koichi Miyata)Fig. 5.6-1 Inundation route at main buildings

Fig. 5.6-2 Restoration of Power (from Unit 6 to Unit 5)

73

5.7
5.7.1

Evaluation of the Released Radioactive Material
Radioactive Material Released into Atmosphere
In order to decrease the assessment uncertainty of radioactive materials

released into the air during the accident, it is preferable to use radiation dose rates
measured at locations closest to the source, such as the stacks. However, as a result ofloss of power, the monitors installed at Fukushima Daiichi plants were unavailable.
Besides, monitoring posts were not available as well, monitoring cars moved around tomonitor dose rate instead. TEPCO made the best use of these data and assessed theamount of radioactive materials that were released into atmosphere. The data are shownin Figure 5.7-1 and Figure 5.7-2.
This assessment was based on the following methods: an inversion analysis
was conducted to calculate the amount of fission products released to trace the peak of
the measured dose rate, and when no peak was observed because of the wind, 1% of the
background dose was used as the contribution from the released fission products.
The calculation results are shown in Table 5.7-1. The total amount of rare gas
and iodine-131 released were both approximately 500 PBq; the amount of caesium-134
and caesium-137 released were around 10 PBq, respectively. The maximum amount offission products was released from Unit 2 after 7:00 on March 15. In addition, fission
products released from the containment venting formed a small fraction of the total
amount, and it was postulated that the fission products released from the venting was
not the main cause of soil contamination in the vicinity.
On March 15, when the fission products were released continuously from Unit
2, a northwest wind, as shown in Figure 5.7-3, persisted and rainfall occurred at thenight in the northwest district, as shown in Figure 5.7-4. Therefore, it is considered thatfission products released from Unit 2 was the main cause of the high contamination in
the district to the northwest of the nuclear power station.
The Nuclear and Industrial Safety Agency (NISA), JAEA, and other agencies
also assessed the amount of radioactive material released into the atmosphere. The
results showed that the amount of caesium released agreed to TEPCO’s result; however,
the amount of iodine released was three times that estimated by TEPCO. TEPCO
explained that the difference may be a result of the assumption that the proportion ofrare gas, iodine and caesium released was constant.

74

5.7.2

Radioactive Material Released into Sea
The assessment of how much fission products was released into sea was basedon the radioactive concentration data measured drainage nearby. The results provided by
TEPCO are given below.
The main fission products sources into sea were contaminated water releasefrom Unit 2 and 3 intake screens, low concentration contaminated water from
centralized waste treatment building, slightly contaminated underground water
accumulated in sub-drain pit of Unit 5 and 6 and wash-out and fall-out fission products
as well.
The total amount of fission products released was approximately 11PBq of
iodine-131, 3.5 PBq of caesium-134, and 3.6 PBq of caesium-137, respectively. Release
rate into the sea decreased dramatically since middle April as shown in Figure 5.7-5.
Table 5.7-2 shows that the total amount of fission products released into sea was lessthan that into air. The assessment results were similar to those given by JAEA andothers.
The detailed information regarding evaluation of radioactive material release isgiven in the reference [5-7].
(Drafting is by Koichi Miyata)Fig. 5.7-1 Diagram of monitoring post arrangement
75

Fig. 5.7-2 Monitoring data
Table 5.7-1 Atmosphere emission results

Fig. 5.7-3 Trace of the plume released at 20:00 on March 1576

Fig. 5.7-4 Rain clouds radar map at 23:00 on March 15

Fig. 5.7-5 Trend of release ratio to ocean
Table 5.7-2 Calculation results of amount of release to ocean (PBq)

77

5.8

Physical Distribution of Resources during the Accident
Since there was insufficient preparedness against long term emergency,
shortage of various essential resources such as batteries, electric power, water injection
facilities, fossil fuel, radiation monitors, radiation protection tools, food, drink, daily
necessities, clothes, bedding, toilets, etc. appeared at the Fukushima Daiichi accident..
5.8.1 On-site Environment
Many roads from Tohoku area to Kanto area were damaged and blocked, andcommunication was worsened by the action of the earthquake and tsunami. Besides, theexplosions in Units 1, 3, and 4 resulted in radioactive contamination around the nuclear
power station. These factors led to delayed delivery and loss of critical resources.
Generally, it takes 3.5 h to reach Fukushima Daiichi from Tokyo, but after theearthquake and tsunami struck and before the release of fission products, it took 7–10 h.
J-village, which is a sports facility located 20 km away from Fukushima
Daiichi, was assigned as the resource transportation center to Fukushima Daiichi and
Fukushima Daini; the radiation level in J-village increased after the explosion at Unit 1
on March 12, and there was a concern that J-village too would have to be included in
the evacuation zone. Hence, TEPCO chose Onahama Coal Center in Iwaki City as thelogistics hub. Aid from outside TEPCO gathered gradually; however, there was a big
confusion in receiving and releasing resources and in warehouse management becauseof reasons such as the following: the Onahama Coal Center was located in the disasterarea; the owners and supplies of many resources were unknown; and the received
resources simply piled up owing to lack of preparation for its reception.
On 15 March, a large amount of fission products was released from Unit 2,
further increasing the radiation level outside the power station.
The drivers who were responsible for delivering resources to Fukushima werefrom companies other than TEPCO. They were not familiar with the local roadconditions, had little knowledge on radiation protection, and did not have the necessary
radiation protection gear. Hence, resource delivery work became difficult. These drivers
were ordered to deliver the resources only to Onahama Coal Center, and then drivers
from Fukushima Daiichi and TEPCO staffs continued delivering these resources to the
nuclear power plants. Several trucks strayed because road information was limited
owing to the unavailability of communication tools. Drivers, who wore full-faceradiation protection masks found it difficult to drive large vehicles on an unfamiliar road.
Difficulties such as the change in the location of the logistics hub, lack of heavymachinery to handle resources, and lack of capable workers delayed the resource78

delivery considerably; many resources did not reach the right destination. Besides,
because the communication condition was poor, many resources were piled up at
unintended locations. In conclusion, the urgently needed resources were not deliveredon time to the nuclear power plants.
5.8.2 Various Resource Supports for Fukushima Daiichi(1) Batteries
Batteries that provide DC power are essential for monitoring, depressurization,and cooling in the event of an accident in a nuclear power plant. Batteries are kept
charged, inspected, and tested periodically, and their performance and function are
maintained; however, Fukushima Daiichi did not have spare batteries. Therefore, since
the evening of March 11, the ERC at the power station scrambled for batteries, and theheadquarters of TEPCO also tried to gather as many batteries as possible regardless of
specifications. The methods for collecting batteries can be largely broken down into
three categories: gathering batteries within the station and purchasing and using
batteries from other TEPCO facilities as shown in Table 5.8-1 and Figure 5.8-1.
(a) Batteries from TEPCO Headquarters
TEPCO headquarters started collecting batteries and chargers soon after they
comprehend the damage of electric equipment. Initially, batteries were collected
regardless of specifications; however, it was noticed that it was impossible to handle
heavy and large batteries at Fukushima Daiichi; hence, car batteries were gatheredinstead. Until the morning of March 12, a thousand 12 V-batteries had been ordered.
Since it took time to get permission to use the highway and to get out of Tokyo, it was
only around 0:00 on 14 March that trucks arrived at Onahama Coal Center. At 21:00 on
March 14, three hundred twenty batteries loaded in two large trucks headed for
Fukushima Daiichi from the hub. Hereafter, the delivery of batteries to the site wasplanned to continue; however, the explosion at Fukushima Daiichi resulted in thesuspension of delivery until March 17.
Additionally, support staff from Kashiwazaki-Kariwa Nuclear Power Stationbought twenty 12 V-batteries in Kashiwazaki City and transported them to Fukushima
Daiichi on March 14.
(b) Batteries from Other TEPCO Facilities
The staff from the head office of TEPCO collected various batteries fromthermal power stations and branch firms and delivered them to Fukushima Daiichi.
79

Workers removed and delivered fifty 2 V-batteries (12.5 kg each) from the siteof Hirono thermal power station, which is close to Fukushima Daiichi, and delivered.
NISA allowed the TEPCO headquarters to use Self-Defense Force’s (SDF’s) helicopter
to deliver batteries, and they arrived at the site at around 1:20 on March 12. The DD-FP
in Unit 1 and reactor water level indicator in Unit 3 were restored using these batteries.
Moreover, 2 V-batteries were gathered from Kawasaki thermal power station
and Tokyo branch firms and were transported; however, owing to their heavy weight,
some of which is weighing about 140 kg, and the lack of heavy machinery at Fukushima
Daiichi site, these batteries were not efficiently used.
(c) Batteries from Fukushima Daiichi
Since the locality had been evacuated, it was impossible to get batteries fromshops in the vicinity of power station; hence, workers decided to purchase batteries from
Iwaki City. At that time, aftershocks still continued, and main roads such as Route 6
were not usable owing to obstacles and sliding stones. Then workers attempted to finddetours and they drove carefully and slowly to avoid uneven and sagging roads.
Furthermore, the numerous evacuation cars caused traffic jams. As Iwaki City was also
affected by the natural disaster, it was not easy to find open stores. As a result, workers
required more than half a day for the round trip, which usually takes only 2 h. Eight 12
V-batteries were purchased and used for the control rooms of Units 3 and 4.
The administration building on the site was also damaged because of theearthquake and frequent aftershocks; hence, access to the building was prohibited forthe moment, and workers could not access the safety box. The money required to buy
batteries was collected from the staff working in the ERC room.
At night on March 11, five 12 V-batteries were taken from buses and company
cars on the site and brought to the control rooms of Units 1 and 2 for restoring reactor
water level indication. To ensure sufficient power supply, twenty additional batteries
were taken from private cars. Ten batteries in a series connection supplied power for
opening SRVs to depressurize the reactors of Units 2 and 3.
(2) Power Supply Cars
An investigation of the on-site and off-site power supply facilities made it clear
to the workers on the site that it was difficult to restore power supply instantly. Hence,
they determined to use power supply cars along with available power distribution
equipment in the power station. Power supply cars were collected from TEPCO, otherelectric utilities, and the SDF as shown in Figure 5.8-2.
80

In total, fifteen high-voltage and low-voltage power supply cars from TEPCO,
and seven cars from other electric utilities and the SDF were gathered. Owing to roaddamage and traffic jam, the power supply cars could not reach Fukushima Daiichi in atimely manner. At 17:50 on March 11, the TEPCO head office attempted to transport
these power supply cars using the SDF’s helicopter, but the cars were too heavy to be
carried by airfreight. On March 12, these cars arrived at Fukushima Daiichi by landroute. After the preparation work for cable route investigation, cable procurement,
removing debris, laying cables and so on, on March 13, a power supply car was partly
connected to the system. However, the subsequent explosion damaged cables and poseda serious challenge to power restoration work. TEPCO’s power supply cars were onlyused because the arrival time was before the restoration.
(3) Fire Engines
Since the SBO made water injection pumps unavailable, fire protection systempiping, which had been prepared for the injection line as an AM measure was used.
Though it was not originally expected as the injection equipment, the line was inconjunction with fire engines that had been furnished after the Chuetsu-oki earthquake.
ERC gathered the on-site and additional off-site fire engines as shown in Figure 5.8-3.
Three fire engines were present in Fukushima Daiichi. One was
non-functional; another was located close to Units 5 and 6, and its availability was
unknown because the debris on the road prevented access. Hence, only one was actuallyavailable. On March 13, the debris was cleared away, and the fire engine from thevicinity of Units 5 and 6 was used for water injection, after confirmation of its
intactness and availability.
In addition, seven fire engines were provided by Minamiyokohama Thermal
Power Station and Kashiwazaki-Kariwa, and five fire engines were obtained from otherelectric utilities, the government, and the SDF. On March 12, these fire engines werearrived at Fukushima Daiichi and were used for transferring fresh water and seawater
into the fire protection tank, and injecting cooling water into the reactor core.
The detailed information regarding the physical distribution of resources isrecorded in the reference [5-1], and the details of this seminar are also given inreferences [5-8].
(Drafter: Atsuhi Kawamoto, Koichi Miyata)81

Fig. 5.8-1 Status of Materials Transportation (Batteries)

82

Table 5.8-1 List of Battery Procurement

83Fig. 5.8-2 Status of Materials Transport (Power Supply Truck)

Fig. 5.8-3 Status of Materials Transport (Fire Engines)84

5.9 Off-site Emergency Response
5.9.1
Radiation around Fukushima Daiichi during the Accident
Fukushima Daiichi is located in an area named Hamadori facing the Pacific.
The breeze circulation in this region follows a particular pattern, as follows. The windblows towards the east during the night, from the land to the Pacific; changes to asoutherly direction in the early morning, and from a southwesterly to westerly direction
during the late morning. Thereafter it blows in a northwesterly direction until theevening.
24 monitoring posts had been established by Fukushima prefectural
government, and all except the post in Ohno, which is located 5 km from Fukushima
Daiichi, were disabled either in relation to the communication cable breaking, or by theloss of power supply, and/or the tsunami wave. However, several portable monitoring
posts were in operation, and these recorded the level of radiation in Fukushima City,
Koriyama City, Shirakawa City, Minami-Aidzu Town, Minami-Soma City, and Iwaki
City, respectively. This information is presented in Figure 5.9-1.
At 17:46 on March 12, an increase in the air dose rate was detected at
Minami-Soma City, which is located 24 km north of the site, and an amount of 20 μSv/h
was then recorded at 21:00. It was evident that the radioactive plume released by theventing of Unit 1, and/or the hydrogen explosion in Unit 1, was moving in the directionof Minami-Soma at that time. After three hours, the radiation level had reduced by half,and on March 13 it measured 4 μSv/h in relation to radioactive nuclides deposited on
the ground. The radioactive plume released on March 12 then moved northwards. The
monitoring post in Onagawa detected a threefold increase in the air dose rate at 20:40
on March 13, at 22:20 on March 13, and at 1:50 on March 14. The peak level recordedon March 13 was around 21 μSv/h, which prompted Onagawa to release a special reportunder Article 10 of Special Act for Nuclear Emergency Preparedness and Response(Nuclear Emergency Act).
The venting of Units 3 began in the morning of March 13, and the hydrogen
explosion occurred in Unit 3 at 11:01 on March 14. However, the monitoring posts
outside the site did not detect the release of radioactive material because they weresituated to the landward side of the plant, and there was an offshore breeze at that time.
In contrast, events in Unit 2 affect the monitoring results significantly. As
described in Section 5.3, the reactor core in Unit 2 was severely damaged in the eveningof March 14. The venting of Unit 2 had not been confirmed, and the PCV pressure was
kept at about 700 kPa [abs] until the following morning. Therefore, it is likely that acertain amount of steam containing radioactive materials was leaked outside PCV85

during this time. The explosion from the neighboring unit had opened the blowout panelon the reactor building in Unit 2, and therefore steam with radioactive materials would
easily have been released from the reactor building to the environment. As a result, theon-site radiation level began to increase. The contaminated air then moved southwards
with the north wind. At 00:00 on March 15, the monitoring post in Iwaki City detected
an increased air dose of 0.57 μSv/h, and the peak in the city rose to 23.72 μSv/h at 04:00.
The plume then continued its movement south, and in Kita-Ibaraki City an increase in
the air dose rate of 0.144 μSv/h was detected at 00:20 on March 15. The maximum dose
rate then reached 5.575 μSv/h at 05:50. At JAEA in Tokai-mura, the change began to be
detected by all monitoring posts at around 01:00 on March 15 and the peak followed ataround 07:00. The influence then spread around the Kanto area up to Shizuoka
Prefecture.
As illustrated in Section 5.3, the D/W pressure in Unit 2 was 730 kPa [abs] at
07:20 on March 15. It then decreased to 155 kPa [abs] at 11:25, at which time thepressure gauge in the S/C was already unreliable. The maximum dose rate recorded at
the Fukoshima Daiichi site was 12 mSv/h, and this was recorded in the southeast area ofthe site at 09:00 on March 15. According to meteorological observations made by anon-site monitoring car, the northeast wind continued to blow from midnight of March
14 to the morning of March 15. By noon, the wind changed towards the southeast andremained blowing in this direction until midnight. As shown in Figure 5.9-1, the change
in the air dose rates increased as follows: Shirakawa City increased to 0.44 μSv/h at
13:00, reached a peak of 5.02 μSv/h at 13:30, slightly decreased later but then amaximum of 7.70 μSv/h was recorded at 22:40; Koriyama City increased to 0.06 μSv/h
at 13:00, then increased rapidly to 8.26 μSv/h, fell later to 3.5 μSv/h later, and at 23:00
started decreasing from 3.54 μSv/h; Fukushima City began increasing from 0.11 μSv/h
at 15:10, increased sharply at 16:30, reached a maximum of 24.24 μSv/h at 18:40, thenstarted decreasing from 23.9 μSv/h at 23:20.
According to the Automated Meteorological Data Acquisition System inFukushima Prefecture, precipitation of 0.5 mm was observed at 17:00 on March 15
when the air dose rate in Fukushima City was at its peak. It then began to reduce slowly,
and it is considered that nuclides were deposited on the ground within rain, and gamma
rays have since been detected on the ground. Rain and snow was widely spread by
midnight, and the radioactive materials were thus deposited from the flow of plume into
the rain and snow. Consequently, a highly contaminated distribution occurred to thenorthwest of Fukushima Daiichi. The Ministry of Culture, Sports, Science andTechnology (MEXT), with the cooperation of U. S. Department of Energy (DOE),
86detected the dose rate within 100 km from Fukushima Daiichi by aircraft monitoring,
and a dosimetry map with accumulated amounts of caesium-134 and caesium-137 on
the ground was obtained. This shows that the distribution was strongly affected by therainfall when the radioactive plume passed through, indicating a non-uniform
distribution.
At 10:50 on March 16, the dose rate in Iwaki City reached a peak of 18.87
μSv/h. On March 18–20 in Minami-Soma City, and on March 21 in Iwaki City, the dose
rate increased as a result of plume passing, but never exceeded 7 μSv/h.
5.9.2

Emergency Response
At 15:42 on March 11, according to Article 10 of Nuclear Emergency Act,TEPCO released a special report to NISA saying that Fukushima Daiichi Units 1–5
were in a station blackout situation. At 16:32, TEPCO recognized that the event
corresponded to the incident described by Article 15 of Nuclear Emergency Act, andthat there was a possible loss of cooling function, and thus reported this to NISA. In
response to this, at 17:45, NISA began the petition process under Article 15, and at
18:22 the Minister of Economy, Trade, and Industry reported the emergency to thePrime Minister. At 19:03, the Prime Minister declared a nuclear emergency situation,and established the Nuclear Emergency Response Headquarters and local response
headquarters.
The evacuation of people living within 2 km of Fukushima Daiichi was firstordered by the Fukushima governor at 20:50 on March 11. The Investigation Committee
on the Accident at the Fukushima Nuclear Power Stations of Tokyo Electric Power
Company (Governmental Investigation Committee) reported that decision to evacuate
within 2 km was based on a usual emergency response exercise. The evacuation wasthen extended to a 3 km radius at 21:23, and those within 10 km were directed to shelter
within their houses. The justification for choosing a 3 km zone was based on theInternational Atomic Energy Agency’s (IAEA’s) recommendation adopted in Japan’s
Nuclear Emergency Guidelines in 2007, that a Precautionary Action Zone (PAZ) shouldbe established around 3~5 km from the site. The pressure of PCV then increased and theventing operation occurred late in Unit 1. At 05:44 on March 12, the evacuation radius
was extended to 10 km. The Governmental Investigation Committee has since reported
that the Emergency Planning Zone (EPZ) was conservatively defined considering theproceedings of the event. However, at 15:36 on March 12 the hydrogen explosion
occurred in the reactor building of Unit 1, and at 18:25 the evacuation radius was
extended to 20 km. It is noted that the choice of 20 km was not related to any specific87authority regulations.
At 11:01 on March 14, the reactor building of Unit 3 exploded. After 06:00 on
March 15, the sound of an explosion was reported near Unit 2 and the reactor buildingof Unit 4 was damaged. Fire then broke out. Given that the multi units were in a severe
condition, government officials dictated that inhabitants living between 20 km and 30
km of the site should also remain sheltering within their houses.
5.9.3

Initial Protective Action regarding Contamination of Foodstuffs and Water
On March 15, high concentrations of radioactive iodine and caesium weredetected in samples of surface mud and grass. Because the area that the plume traversed
remained highly contaminated for quite a long period, NSC suggested gathering an
official emergency team to manage the early monitoring of milk, drinking water, andagricultural crops.
The Ministry of Health, Labor and Welfare (MHLW) has jurisdiction over the
hygiene protection of food and drink on the basis of the Food Sanitation Act. Before theaccident, there was no specific established standard for levels of radioactivity in foodand drink. On March 17, MHLW issued the notice named The Handling of FoodContaminated with Radioactivity. In this notice, a provisional regulation value wasshown according to Indices Relating to Limits for Food and Drink Ingestion, which was
included in the Nuclear Emergency Guideline established by NSC. MHLW notified
local government that food and drink containing above this radioactivity limit should
not be consumed, based on Article 6, No.2 of the Food Sanitation Act. MHLW also
notified that drinking water, and milk and dairy products containing over 100 Bq/kg ofradioactive iodine, should not be used in infant formula nor in milk. Table 5.9-1 shows
the number of samples that exceeded the provisional regulation value until March 31 of2012. The percentage was less than 1 % in Japan. Even in Fukushima Prefecture, only
3.3% of food stuffs exceeded the provisional regulation value.
Since April 1, 2012, the MHLW have indicated a new standard for radioactive
substances found in food. In terms of radioactive caesium, the level of 5 mSv per year
from the previous provisional regulation value was reduced to 1 mSv per year. Thus theradioactive caesium concentration in common food such as vegetables and grain islimited to 100 Bq/kg, which is one fifth of the previous value. Table 5.9-2 shows thenumber of samples that exceeded the new regulation value from April 1, 2012. It can beseen that the number of samples exceeding the new standard in seafood and wild animal
meat is large, but the average percentage is low, so that in all the prefectures it is only
1.2%, and is 4.3% in Fukushima Prefecture.
88

(1) Drinking Water
Radioactive iodine (iodine-131: 177 Bq/kg) was detected in samples takenfrom the water supply in Fukushima City for the first time on March 16. Levels of 308
Bq/kg were then found in Kawamata Town on March 17, exceeding the provisional
regulation value of 300 Bq/kg. In addition, iodine-131 was also detected in Koriyama
City, Shirakawa City, Iwaki City, and in Iitate Village. In Iitate Village, 965 Bq/kg
iodine-131 and 153 Bq/kg iodine-132, which were the maximum value ever detected inrelation to this accident, were detected in tap water on March 20.
On March 19, MHLW informed prefectures that if radioactive substances indrinking water exceeded the provisional regulation value then citizens should: (1)refrain from drinking the water, (2) use it for domestic water, (3) drink the water if there
was no available substitute. On March 21, since radioactive iodine in tap water was over
100 Bq/kg, it was advised that it should not be used for dissolving infant milk powder.
On March 21, the intake of tap water in Iitate Village was restricted, but this restriction
was subsequently lifted on April 1 (except for in relation to infants; this restriction was
finally lifted on May 10). In addition, tap water for use by infants was also restricted ina total of 20 water supply industries distributed in Fukushima, Ibaraki, Chiba, Tochigi,
and Tokyo.
On April 4, MHLW promulgated a notice named The Handling of Radioactive
Materials in Tap Water, that conducted monitoring activities of tap water. It was advisedthat the inspection would be performed at least weekly in the key areas, which included
Fukushima Prefecture and its neighboring 10 prefectures. If the investigation result was
found to be in excess of the regulation value, the investigation was then to be performed
daily. However, all results were below the regulation value after April. In terms ofradioactive iodine, some tap water exceeded the regulation value of 100 Bq/kg in March,and the ratio was the highest from March 16 to March 20. None of tap water had
exceeded the regulation value of radioactive caesium.
(2) Food
The iodine-131 content in raw milk taken from Kawamata Town in Fukushima
Prefecture on March 16 was found to be 1,190 Bq/kg, greatly exceeding the provisional
regulation value (milk, dairy products: 300 Bq/kg). The iodine-131 content in raw milktaken from Mito City and Kawachi Town of Ibaraki Prefecture on March 19 was foundto be 1,700 Bq/kg. In Takahagi City of Ibaraki Prefecture, 15,020 Bq/kg of iodine-131
and 524 Bq/kg of caesium-137 were measured in spinach picked on March 18. In
89

Fukushima Prefecture, the highest contents of iodine-131, caesium-134, andcaesium-137 in spinach taken on March 21 were found to be 19,000 Bq/kg, 20,000
Bq/kg, and 20,000 Bq/kg, respectively. These values exceeded the provisional
regulation value for vegetables, where the value set for radioactive caesium was 500
Bq/kg, and the regulation value of radioactive iodine for vegetables except roots andtubers was set at 2,000 Bq/kg.
In accordance with the provisions of Article 20, Item 3, of the Nuclear
Emergency Act, on March 21, the Prime Minister, who is chief of the Nuclear
Emergency Response Headquarters, ordered the restriction of shipping of raw milk from
Fukushima Prefecture and spinach and kakina (a type of brassica) produced from
Fukushima, Ibaraki, Tochigi, and Gunma prefectures. On March 23, a restriction foringesting vegetables produced in Fukushima Prefecture was issued. The target of theingesting restriction was non-head leafy vegetables (e.g., spinach, cabbage), andcruciferous brassicas (e.g., broccoli, cauliflower). On April 4, the MHFW issued thenotice named The Handling of Provisional Regulations on Radioactive Material in Food.
In the notice, inspection plan conducted by local government was announced. As withthat of tap water, the inspection for food was conducted weekly in Fukushima
Prefecture and its 10 neighboring prefectures. The notice showed the government
guidelines for restriction on shipping and ingestion, and the requirements for eventual
lifting of the said restrictions.
On April 1, 4,080 Bq/kg of iodine-131 was detected in young ammodytidae
fish in Ibaraki Prefecture. However, NSC had not provided Indices Relating to Limits
on Food and Drink Ingestion for fish and shellfish. Therefore, on April 5, the MHFW
announced that the content of iodine in seafood should have the same regulation value
as that of vegetables (2000 Bq/kg). On April 20, a restriction was issued for shippingand ingesting young ammodytidae fish from Fukushima Prefecture.
On April 8, the Nuclear Emergency Response Headquarters issued an approachfor the regulation of rice planting. If a farmland was highly contaminated the planting
would be restricted to prevent the production of rice exceeding the provisional
regulation value. Postulating the transfer coefficient from soil to rice as 0.1, the limit forradioactive caesium in soil was set at 5,000 Bq/kg. On April 22, the chief of the Nuclear
Emergency Response Headquarters instructed the restriction of planting rice togovernors, based on the above approach.
In Fukushima Prefecture, the shipping and/or ingestion of certain food, such as
spinach, mushrooms, and bamboo shoots, is still restricted now (March 22, 2013). In
other prefectures (Ibaraki, Tochigi, Gunma, and Chiba) where shipping restriction on90

milk and vegetables was issued, the restriction had already been lifted. In June of 2011,
caesium-137 exceeding the provisional regulation value was detected in tea leaves, so
the shipping of this item was restricted in many prefectures including Kanagawa. In
July of 2011, caesium-137 exceeding provisional regulation values was found in beef; aresult of contaminated rice straw fed to the animals. A restriction for ingesting beef wastherefore issued in Fukushima, Miyagi, Iwate, and Tochigi prefectures.
5.9.4

Implementation of Additional Protective Actions
On March 15, a high air dose rate was measured throughout Fukushima
Prefecture, in particular northwest of Fukushima Daiichi. MEXT conductedmeasurements using a monitoring car throughout the area at between 20 km and 30 kmof the site, where inhabitants remained sheltered inside house. At around 21:00 on
March 15, an air dose rate as high as 200–300 μSv/h was detected in Namie Town,
located 20 km northwest of Fukushima Daiichi. On March 17, the maximum air dose
rate reached 170 μSv/h at the area called monitoring point No. 32, located 30 km
northwest of Fukushima Daiichi. On March 18, NSC requested NISA to check theexistence of residence and the terrain around the area. Furthermore, NSC requested
MEXT to allocate integrating dosimeters at point No. 32.
On March 25, the NSC published Advice on Emergency Monitoring andProtection Measures. Taking into account the information from previous monitoring
activities and the possibility of release in the future, NSC suggested encouraging
residents living between 20–30 km from the site, who were sheltering in their houses, toevacuate voluntarily. NSC also advised that the area to be protected should be alteredaccording to the monitoring data, although this was not an official requirement at thistime.
On March 27, the Resident Safety Team of the Nuclear Emergency Response
Headquarters told NSC that there were about 200 residents sheltering in houses aroundmonitoring points No. 31 and No. 32 (Tsushima District in Namie Town), and aroundNo. 33 (Warabidaira-Nagadoro District in Iitate Village, where a relatively high dose
rate was observed). On April 6 this information was further revised as 128 residents in
Tsushima and 228 residents in Warabidaira-Nagadoro. The Resident Safety Team thenreported NSC that effective dose of the residents would reach to the index for sheltering
and/or evacuation regarding to Nuclear Emergency Guideline, that was set at 10–50
mSv. NSC replied that a conservative estimation of 15–35 days remained for the dose toreach 50 mSv, which was the index for evacuation. This estimation was based on the air
dose rate at point No. 32 measured from March 17 (170 μSv/h) to March 26 (45 μSv/h).
91

The accumulation dose on March 26 was estimated as being about 15 mSv, assuming 8
h of outside activity and 0.4 as the shielding factor.
On March 30, IAEA advised the Japanese government on the web-site of the
Fukushima Update Log to carefully examine its protective action established for the
residents around the area of Iitate. Their dose analysis was performed based on soil
contamination of iodine-131 and caesium-137. Appling an Operational Intervention
Level (OIL), which was later published in the safety guide GSG-2 in May 2011, a doseof iodine-131 exceeding the limit of OIL was estimated at Iitate. Considering advice
from NSC, the cabinet examined the possibility of extending the evacuation zone, andalso considered changing the protective action for residents located at around 20–30 km
from the site, who were sheltering in their houses.
On April 10, the Nuclear Emergency Response Headquarter asked NSC
whether there should be any alteration of the emergency response action in accordancewith Item 5, Article 10 of the Nuclear Emergency Act. NSC proposed that the area
where the cumulative dose could reach 20 mSv within one year after accident should be
defined as Deliberate Evacuation Area. In 2007, the International Commission ofRadiological Protection (ICRP) recommended that in order to protect the public from anemergency exposure situation, the reference level should be 20–100 mSv (emergent andannual). On the basis of the ICRP’s recommendation and the IAEA’s regulation, NSC
put forward a proposal to use 20 mSv per year as the criteria. Some areas located at a
distance of 20–30 km from Fukushima Daiichi were considered to be beyond the
Deliberate Evacuation Area. However, as the accident situation was considered to beunstable, this area was proposed as being an Emergency Evacuation Preparation Zone.
Based on the proposal of NSC, on April 11 the chief cabinet secretaryannounced the setup of both Deliberate Evacuation Area and Emergency Evacuation
Preparation Zone. On April 22, two zones were formally determined on the basis of
Item 3, Article 20 of the Nuclear Emergency Act. Prior to this formal indication, an
explanation was given to the municipalities. The previous evacuation zone was decreed
a restricted area where access was inhibited, and the sheltering instruction was lifted at a
distance of 20–30 Km from Fukushima Daiichi. The residents living in Deliberate
Evacuation Area should, in principle, be evicted from their homes for one month. The
residents living in Emergency Evacuation Preparation Zone should prepare for eviction
or for sheltering at home. In this area, continuous voluntary evacuation was encouraged.
Children, pregnant women, those who required nursing care, and medical patients were
kept outside this area. Since June 2011, some regions in Minami-Soma City and Date
City where the cumulative dose in one year was considered likely to be higher than 20
92

mSv were declared to be Specific Spots Recommended for Evacuation. Notification, theprovision of information, and evacuation support were provided to the residents.
5.9.5

Estimation of Residents’ Exposure DoseAs described above, during the initial stages of the accident protective actions
were carried out sequentially. However, an estimation of the exposure of individuals
was neither rapid nor adequate.
On March 25, NSC decided that thyroid measurements should be taken from
1,149 children living in Iwaki City, Kawamata Town, and Iitate Village, and this was
performed using an NaI scintillation survey monitor. With the exception of inadequate
measurement affected by the background dose, the result showed that the screening
level in 1080 children was below 0.2 μSv/h, which is equal to an equivalent dose of 100
mSv in the thyroid of one-year-old baby.
Since May 2011, a residents’ health management survey has been implemented
in Fukushima Prefecture. Initially, the survey was implemented for the evacuees ofNamie Town, Iitate Village, and Yamakiya District in Kawamata Town. Later however,a questionnaire pertaining to behavior and a record of diet was collected throughout theentire prefecture. Based on this questionnaire and the monitoring data, the cumulative
effective dose of external exposure over four months (March 11, 2011–July 11, 2011)
was estimated. The return rates of the questionnaire were: 56% for the initial survey,
and 22.9% for the whole prefecture survey. Except for people exposed to radiation in
their working lives, out of 119,450 individuals the highest cumulative effective dose
was found to be 25 mSv, and only 117 individuals had readings higher than 10 mSv.
Approximately 99% of people had measurements of less than 10 mSv until August 31,
2012. The committed effective dose of internal exposure was estimated using a wholebody counter for 81,119 residents from the areas in the Deliberate Evacuation Area andin Futaba County. The result showed that the maximum dose was 3 mSv for two
residents, and less than 1 mSv for 81,093 residents.
5.9.6

Protective Actions in Long-term Contaminated Areas
On May 19, NSC issued Basic Thinking on Radiation Protection, based on theevacuation status and environmental contamination due to the release of radioactive
materials. Subsequently, on July 19, taking the need for various protective actions into
account, Basic Thinking of Radiation Protection Aiming to Lift Evacuation andReconstruction was brought forward. Various systems such as radiation protective
actions in accordance with ICRP’s recommendations in 2007, environment monitoring,93

individual dose estimation, health evaluation, decontamination, and individuals’
participation to radiation protection were discussed and issued. On August 4, at therequest of Nuclear Emergency Response Headquarters, NSC issued the Principles for
Lifting Restriction and Changing Protective Actions. It was considered that before
lifting the evacuation order it was necessary to justify this, and the effect of evacuation
should be compared with other protective actions such as decontamination. It wasconsidered that local government and residents should be involved in the decision
making processes. NSC stated that Exposure less than 20 mSv per year needed to bereached in order to lift the evacuation, and the exposure should be continuously andreasonably reduced for long term in order to reach 1 mSv per year. On September 19,
recovery plans were formulated in all municipalities in Emergency Evacuation
Preparation Zone, and on September 30 the Nuclear Emergency Response Headquarters
lifted the Emergency Evacuation Preparation Zone.
On December 26, when on-site accident mitigation activities had achieved tostep 2, the Nuclear Emergency Response Headquarters reviewed the restricted area
within 20 km of the site and Deliberate Evacuation Area. Three zones were designated.
Restriction was gradually lifted in “zone in preparation for the lifting of the evacuation
order”, where the annual cumulative dose is less than 20 mSv. “Restricted residence are”
is where the annual cumulative dose is more than 20 mSv. “Difficult-to-return zone” iswhere the annual cumulative dose is currently more than 50 mSv, and it would not be
possible to reduce the dose to less than 20 mSv per year even after the period of five
years. Based on these categories, the Nuclear Emergency Response Headquarters
conferred with Fukushima Prefecture, municipalities, and the residents, and on March
30, 2012, issued the regulation accepting the advice from NSC.
(Drafter: Toshimitsu Homma)

94

Fig. 5.9-1 Change in the air dose rate at seven places within Fukushima prefecture

Table 5.9-1 Food exceeding the provisional regulation value (until March 31, 2012)
Fukushima Prefecture
Type of Food

All

Number of

Exceeding

Ratio

Number of

Exceeding

Ratio

SampleRestriction(%)

SampleRestriction(%)

Vegetables

8302

302

3.6

21089

451

2.1

Seafood

3574

227

6.4

9183

245

2.7

Milk・Dairy products

685

18

2.6

2961

23

0.8

Meat・Egg

6789

165

2.4

93223

286

0.3

Grains

1860

2

0.1

5552

2

0.0

Others

256

4

1.6

3791

197

5.2

Total

21466

718

3.3

135799

1204

0.9

95

Table 5.9-2 Food exceeding the new regulation value (between April 1, 2012 andNovember 7, 2012)
Fukushima Prefecture
Type of Food

All

NumberExceeding

Ratio

NumberExceeding

Ratio

of Sample

Restriction(%)

of Sample

Restriction(%)

Agricultural products

9271

139

1.5

34408

550

1.6

Stock farm products

5754

1

0.0

89393

3

0.0

Seafood

3926

609

15.5

12686

793

6.3

Milk・modified milk powder

304

0

0.0

3129

0

0.0

Meat of wild animals

161

96

59.6

638

199

31.2

Drinking water

18

0

0.0

1326

13

1.0

Others

1432

60

4.2

5904

0

0.0

Total

20866

905

4.3

147484

1704

1.2

96

5.10
5.10.1INES
INES Introduction
The International Nuclear and Radiological Event Scale (INES) is used forprompt communication to the public the safety significance of events which occur
relating to nuclear or radiation utilization using the internationally common scale. Whenan event has higher significance than the reporting criterion, the rated scale as well as
the description of the event is notified to IAEA.
Events are classified into levels as below: events without safety significanceare classified Below Scale or Level 0. Levels 1–3 are “incidents”, and Levels 4–7 are“accidents”. These levels are decided based on three areas of impact: “people and theenvironment” is based on the doses received by the public or workers; “radiologicalbarriers and control” is based on significant actual failures of radiological barriers; and“defense in depth” is based on the potential consequences of events and rated with the
behaviour of the safety function, frequency of the initiator, and so on. The “INES user’s
manual” was prepared for the detail assessment. The latest manual was issued in 2008.
At least one National Officer (NO) is designated to each INES member state.
When the regulatory authority receives event information from licensees, it rates theevent in accordance with INES user manual and declares the INES scale of the event
immediately. The significance of the event is instantly informed like “this event is level
XX on the INES scale” to the public. This description is similar to the magnitude ofearthquake. The NO also reports the event to IAEA if the rated level of the event is 2 orhigher. Responsible officers for INES in IAEA inform the event descriptions and ratings
to NOs of all member states through the NEWS system, which is an information systemmanaged by IAEA.
As described above, immediate rating, or provisional rating just after the event
is carried out by the NO. In the later stage, the adequacy of the provisional rating is
examined by the experts committee that determines the final rating.
More detailed information regarding INES can be found in references 5-9(English) and 5-10 (Japanese).
5.10.2

INES Rating of Fukushima Accident
NISA carried out INES rating four times regarding the Fukushima Daiichi
accident. Each rating was provisional and was based on “what were believed to occur
with high confidence” at the time of the rating.
The first provisional rating was carried out and published around 0:30 on
97March 12 5-11). The events at Fukushima Daiichi Units 1, 2, and 3 were rated Level 3
based on the “defense in depth” criterion owing to loss of heat removal. The first rating
was carried out also for the events at Fukushima Daini Units 1, 2, and 4, and they werealso rated level 3 owing to the same reason.
The second provisional rating was carried out in the evening on March 12 5-12).
The event at Fukushima Daiichi Unit 1 was re-rated Level 4 with the “radiologicalbarriers and controls” criterion. This is because the radiation level at the site boundary
was increased at 16:17 on March 12, and it was believed to be caused by the Unit 1
event.
The third provisional rating was carried out on March 18 5-13). The events at
Units 1, 2, and 3 were rated level 5 on the basis of the “radiological barriers andcontrols” criterion. This is because core melt was believed to have occurred based on
the following observations made at that time: (1) there was a certain period of time
where the reactor water level was lower than the top of active fuel; hence, the fuel was
supposed to be at a high temperature; (2) explosions that were possibly caused byhydrogen burning had occurred; (3) The radiation levels inside and outside of the
facility were increased.
The fourth provisional rating was made on April 12 5-14) based on the “people
and the environment” criterion.
In the rating with this criterion, the radiological dose to the public and theworkers as well as the quantity of radioactive material released to the environment istaken into account. This criterion is prepared on the basis of the following discussion
5-9): The simplest approach to rating actual consequences to people would be to base therating on the doses received. However, for accidents, this may not be an appropriate
measure to address the full range of consequences. For example, the efficient
application of emergency arrangements for evacuation of the public may result in
relatively small doses, despite a serious accident at an installation. To rate such an event
purely on the doses received does not indicate the true significance of what happened at
the installation, nor does it take account the potential widespread contamination. Thus,
for the accident levels of INES, which is level 4-7, criteria have been developed basedon the quantity of radioactive material released, rather than the dose received.
In April, NISA estimated the total amount of radioactive material, equivalently
converted into iodine-131, released from all the units at Fukushima Daiichi as 3.7×1017
98

Bq, based on the severe accident analyses carried out by JNES computationally. On theother hand, NSC estimated it as 6.3×1017 Bq based on the reverse calculation of acomputer code for environmental radioactivity transport analysis, using the monitoring
results. Although both values have large uncertainties and are significantly different
from each other, they exceed the criterion of INES level 7, i.e., “an environmental
release corresponding to a quantity of radioactivity radiologically equivalent to a releaseto the atmosphere of more than several tens of thousands of terabecquerels (10 17 Bq) of
131
I”. Therefore, NISA concluded that the rating of the Fukushima Daiichi accident islevel 7. This rating was applied to the entire site, in contrast to previous ratings thatwere applied to each unit separately. Further, TEPCO stated on April 21 that the total
amount of iodine-131 and caesium-134 and -137 discharged into the sea was around
4.7×1015 Bq.
As of March 2013, the INES rating is still provisional. However, according to
the analysis results of NISA, NSC, TEPCO, and other research organizations, the ratinglevel of this accident did not change and has remained at level 7.
5.10.3Issues in the INES Rating and International Correspondence
The Fukushima Daiichi accident was the first Level 7 event that occurred since
the INES was developed in 1990. The accident was rated during the emergency
situation when the accident was still progressing and not enough information wasavailable. Such a situation had not been experienced before.
There are many issues that need to be addressed regarding INES rating in lightof the Fukushima Daiichi accident. They are described below:- INES aims to promptly communicate to the public the safety significance of
nuclear and radiological events. When plant parameters and radiation
monitoring fail, it is difficult to understand accident progression and therelease and transport of radioactivity.
- As a result, level 5 rating was made one week after the accident initiation,-and the level 7 rating was made one month later. This is far from “prompt
communication”.
The INES ratings were carried out on the basis of “what believed to occur
with high confidence” at the time of the rating, as mentioned. According to
the accident progression as well as information obtained, however, therating was changed from lower to higher levels. There was a criticism that
the INES rating led the public to feel the accident less severe than in reality.
99

-

-

The amount of radioactive material released to the environment wasapproximately 10% of that released during the 1986 Chernobyl accident,
which is the only other nuclear accident to have been rated as a level 7 event.
Hence, the question was raised as to why the Fukushima Daiichi accident
was also rated as level 7.
INES ratings were always carried out for each radiation source unit (single
reactor in the case of nuclear power stations) before the Fukushima Daiichi
accident. In the level 7 rating of the accident, however, there were releasesof radioactive material from the multiple units and it was difficult toevaluate the amount of release from respective units; hence, the rating was
made for the site. It is debatable how INES should be applied to suchsituation.

There is concern regarding the adequacy of NISA’s evaluation of the source
terms, which is the amount of radioactive material released to the environment. As
described above, NISA estimated the source terms by referring to JNES’s calculation
with severe accident analysis codes. At the time of the calculation, however, it wasalmost impossible to carry out an accurate evaluation of the source terms from eachunit.
According to TEPCO’s assessment result in Clause 5.7.1, the most dominant
contributor to the total source terms from the site is believed to be the source terms from
Unit 2 after 7:00 on March 15, and this was through the DW.
While during the accident and even long after the accident, however, most of
the related people believed that the leakage was through the S/C in Unit 2. As describedin Section 7.7, it was reported that an explosion was heard at 6:00~6:10 on March 15,
close to the S/C in Unit 2. It is apparent that this sound was caused by hydrogen
explosion in the reactor building of Unit 4. Almost simultaneously, the S/C pressure in
Unit 2 decreased. This is believed to be an instrument failure. The coincidence of thesound and the pressure decrease led to the misunderstanding that the S/C of Unit 2 had
failed.
As generally known, there is a considerable difference between the analysed
results obtained by assuming leakage through the DW and that through the S/C.
Regardless of whether the scrubbing effect of the suppression pool is taken into
consideration, there was a difference of approximately two orders of magnitude between
the calculated source terms. In the source term evaluation for the INES rating evaluation,
the source terms should be evaluated based on the environment monitoring result, while100the calculation results with severe accident analysis codes should be treated only as
reference.
After the Fukushima Daiichi accident, the international society was concerned
whether the INES can play its original role of communicating to public underemergency situations. In June 2011, at the Fukushima Ministerial Conference on
Nuclear Safety, IAEA Director General Yukiya Amano entrusted to the INES Advisory
Committee (INES-AC) to consider ways of making the INES more effective as acommunication and information tool.
INES-AC had worked on this issue for one year. In the AC meetings, firstly,“INES aims for whom” was discussed. In conclusion, INES cannot play its part as acommunication tool to people directly affected by the emergency. However, INES canbe a useful tool to communicate the significance of the accident to the people outside ofthe emergency.
INES-AC developed an additional guidance that provides a method to carry out
INES rating and to communicate to the public during progressing events. It was
attached to the ‘Use of INES’ guideline that had already been drafted by AC. The
additional guidance states the following:- INES rating should be carried out if some serious situations such as coremelting are believed to have occurred with a high confidence.
- In the case the situation is still evolving, however, care should be taken
about the method of communication.
-

Carrying out unnecessary repetitive INES ratings for a single event should
be avoided.
Furthermore, concerning the question “why the Fukushima Daiichi accident
was rated on Level 7 that was as same as Chernobyl accident”, it was concluded that the
rating for the Fukushima Daiichi accident remains as Level 7. Because Level 7 means“very serious accident”, it is not appropriate to subdivide it.
(Drafter: Kiyoharu Abe, Naoto Sekimura)101

Reference

5-1

Tokyo Electric Power Company, “Fukushima Nuclear Accidents Investigation
Report”, June 20, 2012.
http://www.tepco.co.jp/en/press/corp-com/release/2012/1205638_1870.html

5-2

K. Miyata, “The Accident at Fukushima Daiichi Unit 1”, Lecture slides for the
2nd Seminar, May 8, 2012, in Japanese.
Tokyo Electric Power Company, “Estimation of Condition of Reactor Cores
and the Primary Containment Vessels”, March 12, 2012, in Japanese.
http://www.tepco.co.jp/nu/fukushima-np/images/handouts_120312_02-j.pdf
K. Miyata, “The Accident at Fukushima Daiichi Unit 2 and Unit 3”, Lecture

5-3

5-4
5-5

5-6
5-7

5-8

5-9

5-10

5-11
5-12
5-13

slides for the 3nd Seminar, June 26, 2012, in Japanese.
K. Miyata, “The Fukushima Daiichi Accident, mainly regarding to the Spent
Fuel Pool in Unit 4”, Lecture slides for the 4nd Seminar, August 10, 2012, inJapanese.
K. Miyata, “The Accident at Fukushima Daiichi Unit 5 and Unit 6”, Lectureslides for the 7nd Seminar, November 2, 2012, in Japanese.
Tokyo Electric Power Company, “The Estimated Amount of Radioactive
Materials Released into the Air and the Ocean Caused by Fukushima Daiichi
Nuclear Power Station Accident Due to the Tohoku-Chihou-Taiheiyou-Oki
Earthquake”, May 24, 2012.
http://www.tepco.co.jp/en/press/corp-com/release/2012/1204659_1870.html
A. Kawamoto, “Physical Distribution of Resources at the Fukushima Daiichi
Accident, and the Points for the Improvement of the Logistics”, Lecture slides
for the 7nd Seminar, November 2, 2012, in Japanese.
The International Atomic Energy Agency, “INES, The International Nuclear
and Radiological Event Scale User’s Manual 2008 Edition”, IAEA-INES-2009,
May 2009.
K. Abe, M. Yagi, “INES has been Communicationg the Importance of Events
to Public : Twenty Year Anniversary of International Nuclear and Radiological
Event Scale”, Journal of the Atomic Energy Society of Japan 53(5), p. 344-348,
May 2011, in Japanese.
Nuclear and Industrial Safety Agency, Press Conference, March 12, 2011.
Nuclear and Industrial Safety Agency, Press Conference, March 12, 2011.
Nuclear and Industrial Safety Agency, “INES Ratings on the Events inFukushima Dai-ichi NPS and Fukushima Dai-ni NPS by the Tohoku Regional
Pacific Ocean Offshore Earthquake”, News Release, March 18, 2011.
102

5-14

http://www.nsr.go.jp/archive/nisa/english/files/en20110319-6.pdf
Nuclear and Industrial Safety Agency, “INES (the International Nuclear andRadiological Event Scale) Rating on the Events in Fukushima Dai-ichi Nuclear
Power Station by the Tohoku District - off the Pacific Ocean Earthquake” Press
Release, March 12, 2011.
http://www.nsr.go.jp/archive/nisa/english/files/en20110412-4.pdf

103

6.
6.1

Summary of Other Nuclear Power Plants
Fukushima Daini Nuclear Power Station
Fukushima Daini Nuclear Power Station of Tokyo Electric Power Company
extends over Naraha Town and Tomioka Town of Futaba County in Fukushima
Prefecture. It has four BWRs, whose electrical outputs are all 1,100MW. These plants
began commercial operation between 1982 and 1987, and Unit 1 is of type BWR-5 with
MARK-II containment, while Units 2~4 are BWR-5 with improved MARK-II
containment.
When the earthquake occurred, a seismic intensity 6-upper was recorded in
Naraha Town and Tomioka Town. The maximum horizontal acceleration on the basemat was 277 gal in Unit 3, and the maximum vertical acceleration was 305 gal in Unit 1
as shown in Table 6.1-1. Both these values were below the response corresponding totheir design basis ground motion.
The subsequent tsunami ran-up and reached 15 m around the south part of Unit
1 and the ERC building. Seawater heat exchange building located 4 m above sea level
was badly damaged as shown in Figure 6.1-1. Except for Unit 3, the seawater pumps in
the other three units lost their function, and Fukushima Daini’s ERC released a special
report under Article 10 of Nuclear Emergency Act regarding to the declared emergency
for the loss of the residual heat removal function. Hereafter, the temperature in the S/Cgradually increased to 100 °C and the ERC declared loss of the pressure suppressionfunction based on the Article 15 of Nuclear Emergency Act.
Although the major tsunami warning was being issued, the ERC staff
confirmed the damage of facilities through a walk down. They then started to organize
logistics towards restoration, and equipment was procured through a tie-up with theheadquarters. TEPCO workers in cooperation with partner companies replaced thewater pump motor and lay temporary cables. Four days after the earthquake struck, a
cold shutdown of all units was achieved at 7:15 on March 15.
6.1.1

Outline of the Earthquake and Associated Tsunami
Tohoku-Pacific Ocean Earthquake occurred at 14:46 on March 11, 2011, andsubsequently, at 15:30 a tsunami struck Fukushima Daini. Off the coast of Fukushima
Daini, the height of the tsunami was estimated to be 9 m. An investigation of thetsunami trace at Fukushima Daini showed that the seaside area at Onahama Point (O.P.)+4 m was flooded completely with an inundation height around O.P. +7 m as shown in
Figure 6.1-2.
In addition, the tsunami ran-up along a road at the south-east part of the site104

towards the ERC building. It reached to O.P. +12 m height, and then flowed from Unit 1
area to Unit 4 area. As a result, the inundation depth in the south part of Unit 1 was deep,
Units 2 and 3 were slightly flooded, and no flooding was observed in Unit 4.
The inversion analysis of the tsunami by TEPCO shows that Fukushima
Daiichi was struck by a higher tsunami of 13 m due to the overlapping of three waves as
illustrated in Figure 6.1-3 than the tsunami height of 9 m at Fukushima Daini.
6.1.2
Influence of the Earthquake and Tsunami(1) Effect of the Earthquake
A walk down was carried out immediately after the earthquake occurred. No
abnormality was observed at important seismic safety facilities, but some damage toequipment of seismic classes B and C such as fresh water tanks was noted. In addition,
the seismic response analysis for typical equipment was performed based on the actual
ground motion. The seismic loads on various parts were confirmed to be lower than thatof design basis loads and the stress generated in the piping system was small enough.
Because the circuit breaker was damaged by the earthquake at Shin-Fukushima
Transformer Station, Tomioka No.2 Line (500 kV) became unavailable. In addition,
Iwaido No.2 Line (66 kV) was unavailable due to damage to the thunder arrester. Before
the earthquake, Iwaido No.1 Line was under inspection, so only Tomioka No.1 Line wasavailable out of four lines. Though EDGs are expected to provide power in the large
seismic event, the offsite power also played important role in restoration activities at
Fukushima Daini.
(2) Effect of Tsunami
Except the south building of Unit 3, seawater heat exchanger buildings werestruck by the tsunami, and seawater entered through the equipment hatches. The doors
were designed to open outwards but were broken inwards by the force of tsunami wave.
The RHR cooling water (RHRC) pump, emergency equipment cooling water (EECW)
pump, RHR seawater (RHRS) pump, and the power panels of these equipment, whichwere located on the first floor, were flooded. The pumps of RHRC and EECW werehorizontal pumps, while the RHRS pumps were vertical pumps. Because the positionsof the motors in the horizontal pumps were relatively lower to the ground, the motors of
RHRC and EECW were also flooded and they lost function.
The heat exchanger building (system B) in Unit 3 was not flooded and thefacilities inside were preserved. Therefore, only Unit 3 was shut down by normal
procedures. System B of the EECW pumps of Units 2 and 4, which were located on the105second floor, survived naturally, while system A located on the first floor lost function.
Two RHRS pumps of system B of Units 2 and 4 were installed side-by-side, but only
one remained intact.
The seawater flowed through the equipment hatches and air-conditioner ducts,
submerging the basement. The water that inundated the seawater heat exchanger
building of Unit 1 and 3, flowed through concrete trench to the turbine building of Unit
3, and as a result, the basement of the turbine building was also submerged.
In addition, the tsunami ran-up along the road of the southeast side of the mainbuildings reaching up to the waste treatment building, the ERC building, and the reactor
building of Unit 1. The doors of the waste treatment building and the ERC building
were broken and seawater entered into these buildings. In the reactor building of Unit 1,
the water flowed through the air supply louvers used for ventilation and reached the
EDG through the blower of the EDG.
In terms of the emergency power panels in the reactor building, only one ofthem in Unit 1 was damaged. Further, EDG for system-B of Unit 3, EDG for thehigh-pressure core spray (HPCS) of Unit 3, and EDG for HPCS of Unit 4 wereoperating because the corresponding seawater cooling pumps were not affected by thetsunami.
6.1.3
Response After the Tsunami(1) Reactor Cooling Water Injection
At 14:46 on March 11, when the earthquake occurred, Units 1 to 4, which had
been working at rated power, were shut down by a scram signal. Around 30 minutes
later, due to the tsunami, emergency cooling pumps could not be initiated, and the RHR
system lost function except for system B of Unit 3. At 18:33, this situation was judged
as ‘loss of reactor heat removal function’, which is described in Article 10 of the
Nuclear Emergency Act. Unit 3 was cold-shutdown at 12:15 on March 12 via the RHR(system B) that was not flooded.
In regard to Units 1, 2, and 4, operators followed an emergency operation
procedure (EOP). They closed manually the main steam isolation valve and used SRV todepressurize the reactor pressure. RCIC system was used to inject water into the reactor
core at first. After the RCIC turbine was isolated automatically due to the shortage of
steam from the reactor, water was injected using MUWC, which was prepared as part ofaccident management strategies. Thus, the reactor water level was maintained though
the ultimate heat sink was lost.
106

(2) Containment Vessel Cooling
During water injection to the core, as the RCIC was in operation and the SRV
was opened, the water temperature in the S/C increased to more than 100 °C. Thus, at
06:07 on March 12, it was announced that three units lost function of pressure
suppression, which is the event referred in Article 15 of Nuclear Emergency Act. In
accordance with the EOP, operators initiated containment vessel spray via the MUWC
system. They also used cooling water discharge line of the combustible gas control
system, of which the procedure had not existed in advance, to inject water into S/C. The
temperature and pressure in the PCVs were brought under control as shown in Figure
6.1-4, and operators gained time to restore the residual heat removal function.
In case the residual heat removal function could not be maintained, workers
started to set up the line for containment venting. The so-called “feed and breed” line
was completed except for one action of opening an isolation valve at the S/C side.
However, the cooling of S/C was initiated using the restored RHR before the PCV
pressure reached a determined point of containment venting, and venting was not
executed.
(3) Restoration
After the earthquake, the plant manager immediately organized ERC in theseismically isolated building for dispatching information, appealing for aid, andaddressing restoration activities. In parallel to the operator’s action of maintaining corecooling, the ERC prioritized restoration tasks based on the understanding of the damage
situation of the facilities.
However, walking around was very dangerous because frequent aftershocks
occurred, major tsunami warning continued to be issued, no lighting was available, andthere were many debris and holes all over the site. The tsunami disabled the broadcast
system and communication such as PHS in buildings, and hence the recovery team
could not be dispatched to the site. Around 22:00 on March 11, the recovery team, after
ensuring staffs for communication and safety gears, started to inspect and identify thedamage of the site including the seawater heat exchanger building near the sea.
On the basis of the report from the recovery team, the ERC aimed to restore theheat removal function quickly as well as the pumps in the seawater heat exchanger
building where the damage was small. The inspection and restoration of the flooded
pumps was conducted and wetted motors were replaced. Because the power panels thatprovided electricity to these motors were submerged and had lost function, workers
connected these motors with the power from the waste treatment building, the seawater107

heat exchanger building of Unit 3, and high-voltage power supply cars.
(4) Emergent Transportation of Resources
In order to restore the facility, Fukushima Daini asked the TEPCO headquarters
and the Kashiwazaki-Kariwa Nuclear Power Station to deliver the necessary resources
such as a motor, high-voltage power supply car, mobile transformer, and cable. TEPCO
headquarters and Kashiwazaki-Kariwa confirmed inventories with relevant parties andplanned to transport these necessary resources to Fukushima Daini by all possible routes
such as by air and road.
By around 06:00 on March 13, these resources were delivered at Fukushima
Daini. The motor of EECW(B) and RHRC(D) for Unit 1 were delivered first by an SDF
helicopter from Mie Factory of Toshiba Corporation to Fukushima Airport and then to
Fukushima Daini by land. The motor of RHRC (B) for Unit 4 was transported by truck
from Kashiwazaki-Kariwa by a partner company. Further, the high-voltage power
supply car for EECW (B) in Units 1 and 4 was delivered from branch sections of
TEPCO. In addition, total 9 km long cables were procured from warehouses of TEPCO
and other cooperative enterprises.
The major difficulties during the process of transporting resources by land aresummarized as below: Firstly, many roads were damaged by earthquake and manyroutes were unusable. It was important to determine the possible transit routes to thefacility and to ensure the communication tools that allowed drivers to have thisinformation. Furthermore, the most reliable way was to not only set up a transition
center for resource convergence but to also organize a professional transport team.
Secondly, with respect to the area that might be contaminated by radioactive materials,
it was necessary to inform the drivers ahead of time and obtain prior permission to passthrough this area. In fact, many carriers were held up by police and only staff from thepower station who possessed license for large vehicles were allow to drive trucks to thesite smoothly. This required a transportation centre and a dedicated team in order todeliver reliably without using external carriers.
(5) Restoration of the Heat Removal Function and Accomplishment of Cold
Shutdown
In order to restore the emergency equipment cooling system in the heatexchanger building, access routes were first secured. Workers operated heavy vehicles
to remove the debris and flotsam and to cover the road with gravel where asphalt had
been washed out.
108

Next, workers started to replace the motor of the emergency apparatus coolingsystem pump and laid the temporary cable. The approximately 9 km long temporary
cable was laid at 23:30 on March 13 by 200 workers, as well those from Fukushima
Daini, the Power Distribution division in TEPCO, and partner companies.
Meanwhile, workers were checking the mechanical status of the pump andinstalling the motor. Starting with Unit 1 at 20:17 on March 13, the pump was initiated.
At 1:24 on March 14, RHR pump (B) in Unit 1 was initiated and at 15:42,
RHR pump (B) in Unit 4 started functioning. Then, the site manager declared that theheat removal functions were restored in all units, so that Fukushima Daini was nolonger under the situation defined by the Article 10 of Nuclear Emergency Act. In
addition, workers set up the circulation line that injected water to the reactor core fromthe S/C through RHR heat exchanger (B) using RHR pump (B), and then the water fromthe reactor went through the SRV to return to the S/C. S/C water temperatures decreased
to less than 100 °C in Unit 1 and finally Unit 4 at 17:00 on March 14 and at 7:15 on
March 15, respectively. At this point, the site manager declared recovery from the lossof pressure suppression function in all units, so that Fukushima Daini was no longer
under the situation defined by the Article 15 of Nuclear Emergency Act.
(6) Supports for Achieving the Cold Shutdown
As described above, with support from various agencies, all the units inFukushima Daini achieved cold shutdown. The seismically isolated ERC building
particularly played an important role as a center of restoration work for achieving cold
shutdown. The ERC building was newly built based on lessons learned from the
Chuetsu-oki earthquake, and was in service on July 22, 2010 (8 months before thisearthquake). Its seismic isolation function, communication function, and storing
function greatly enhanced the efficient and effective response to the accident.
About 400 TEPCO staff contributed to ERC and engaged in the accident
response activity under very serious environments without enough food and sanitation
for about 10 days. Many were local staff. They were not even aware of the condition of
their families and houses. They took nuclear safety as the highest priority and devoted
to the restoration work. Without their strong professional spirit, the cold shutdown work
would have been very difficult to accomplish.
The detailed information regarding Fukushima Daini is recorded in reference[6-1], and the details of this seminar are given in references [6-2].
(Drafter: Kazuyuki Nagasawa)109

Table 6.1-1 Comparison between DBE Ground Motion and the Observed Records

Fig. 6.1-1 2F-1 Site Level and Tsunami Run up Height110

Fig. 6.1-2 Tsunami run up height at Fukushima DainiFig. 6.1-3 Differences of the Tsunamis at Fukushima Daiichi and Fukushima Daini111

Fig. 6.1-4 Containment Vessel Pressure (Unit 1)

112

6.2

Onagawa Nuclear Power Station
The Onagawa Nuclear Power Station (Onagawa) of the Tohoku Electric Power
Company extends over Onagawa Town and Ishinomaki City in Miyagi Prefecture. The
station has three BWR. Unit 1 is a BWR-4 with a Mark-I containment vessel, and an
generating capacity of 524 MW. Units 2 and Unit 3 are BWR-5 with improved Mark-I
containment vessels, each with a generating capacity of 825 MW. These units firststarted operation between 1984 and 2002.
Upon the occurrence of the Tohoku-Pacific Ocean Earthquake at 14:46 on
March 11, 2011, Units 1 and 3 were operating normally, while Unit 2 was starting up.
All three units were immediately stopped through the application of the scram
procedure.
After the earthquake, one of five incoming power lines was still operational, as
were all the EDG also worked. The subsequent tsunami did not reach the main buildingsof the facility. The cooling functions of the reactor core and spent fuel pool weresecured. The reactors were successfully brought to a cold shutdown, such that the safetyof the facility was assured.
Even though Onagawa was the closest nuclear facility to the hypocenter, thesafety functions for stopping and then cooling the reactors, and those for containing theradioactive material, all operated as intended. No radioactive material was released into
the environment.
This section addresses the situation and damage at the Onagawa after theearthquake and tsunami, and examines the operation and efficacy of the safety functions.
Detailed information on the Onagawa is provided in the reference [6-3], and a summary
was given in the Seminar [6-4].
6.2.1
Summary of Earthquake and Tsunami, and Subsequent Responses(1) Summary of Earthquake and Tsunami
Tohoku-Pacific Ocean Earthquake occurred at 14:46 on March 11. The huge
earthquake of magnitude 9.0 affected Onagawa, located 125 km from the hypocenter,and 123 km from the epicenter. Six lower on Japanese seismic intensity scale wasrecorded in Onagawa Town. At the sub-basement of the reactor building in Onagawa
Unit 1, 567.5 gal of maximum seismic acceleration was measured; in contrast theprevious maximum record was 251.2 gal on August 16, 2005. Subsequently, at 15:29
the largest tsunami wave struck the site, whose height was about 13 m as indicated by
the tide gauge.
113

a. Examination of Seismic Records
The maximum acceleration observed in the reactor buildings of Onagawa Units
1, 2 and 3 was slightly more than the maximum response acceleration that was based on
basic earthquake ground motion Ss that was formulated after the revision of the Review
Guide for Seismic Design in September, 2006, as listed in Table 6.2-1.
The response spectrum of the seismic records for the basement floor slightly
exceeded the response spectrum of basic earthquake ground motion Ss. Overall,however, the two spectrums were almost equal.
A seismic response analysis was performed based on the seismic records. The
deformation and shearing force of the earthquake-resistant walls of the reactor buildingsof Units 1, 2 and 3 were evaluated, and it was confirmed that the function of the reactor
buildings had not been compromised.

Table 6.2-1 Comparison of maximum acceleration based on observed records andmaximum response acceleration with respect to reference ground motion Ss
Measurement
Location

Maximum Acceleration relative to

Maximum Acceleration (Gal)S-N※

basic earthquake ground motion
Ss (Gal)E-W

Vertical

S-N

E-W

VerticalTop of Roof

2000

1636

1389

2202

2200

1388

Operation Floor (5F)

1303

998

1183

1281

1443

1061

Ground Floor

573

574

510

660

717

527

Basement

540

587

439

532

529

451

Top of Roof

1755

1617

1093

3023

2634

1091

Operation Floor (3F)

1270

830

743

1220

1110

968

Ground Floor

605

569

330

724

658

768

Basement

607

461

389

594

572

490

Top of Roof

1868

1578

1004

2258

2342

1064

Operation Floor (3F)

956

917

888

1201

1200

938

Ground Floor

657

692

547

792

872

777

Basement

573

458

321

512

497

476

Unit 1

Unit 2

Unit 3※ This value is not valid, because the upper limit of the seismic recorder was 2000
Gal.

114

b. Tsunami
The height of the tsunami as observed by the Onagawa tide gauge was about
O.P. +13 m. The effects of the tsunami could be observed in the seaward part of the site.
But the height of the site of the main buildings (O.P. +13.8 m) was higher than that of
the tsunami.
Note: O.P. is the datum level of Onagawa port, corresponding to -0.74 m relative tothe standard mean sea-level of Tokyo Bay. However, given the results of aninvestigation undertaken by the Geographical Survey Institute, the height of thetsunami and the elevation of the site described in this report are those subsequent to
the diastrophism caused by the earthquake. That is, even though the elevation of themain buildings of the Onagawa had been O.P. +14.8 m, in this report their elevation
is described as O. P. +13.8 m.
(2) Response of Each Unit
a. Unit 1
At 14:46, in response to the strong shaking of the earthquake, the reactor was
stopped automatically, through the application of the scram procedure. All of the control
rods were inserted, and the reactor was confirmed as having become subcritical at 15:05.
The external power supply was secured immediately after the earthquake. However, ashort-circuit and ground fault occurred in a high-voltage power distribution board,
namely, the metal-clad switch gear (M/C) of the regular system, leading to the failure of
the start-up transformer that should have received the external power. Consequently, thepumps of the condensate and feedwater systems shut down. In response to this, thereactor was isolated, RCIC system was started to provide coolant to the reactor core,
and the reactor pressure was controlled by the SRVs. Later, the reactor was cooled bymeans of RHR and it ultimately entered the cold shutdown status at 0:58 on March 12.
The plant parameters and notable events are illustrated in Figure 6.2-1.
b. Unit 2
The reactor was starting up when the earthquake struck, but was automatically
stopped through the application of the scram procedure. Since the reactor was
subcritical and the temperature was lower than 100 oC, Unit 2 entered the cold
shutdown status as soon as the operator moved the reactor mode switch to the“shutdown” position at 14:49 on March 11.
115

c. Unit 3
In the same way as for Unit 1, the scram procedure was applied, the control
rods were inserted normally, and the reactor stopped automatically. Subcriticality wasconfirmed at 14:57. The condensate and feedwater systems were initially used to feedcoolant into the reactor core. Since one of the pumps of the turbine auxiliary coolingwater system stopped as a result of being swamped by the tsunami, the operator stopped
the feedwater pump manually. Instead, coolant was provided through the RCIC system,and the reactor pressure was controlled with the SRV. Later, the reactor was cooled bymeans of RHR, and then entered the cold shutdown status at 1:17 on March 12. The
plant parameters and notable events are illustrated in Figure 6.2-2.

Fig. 6.2-1 Reactor core water level and water temperature in Unit 1

Fig. 6.2-2 Reactor core water level and water temperature in Unit 3116

6.2.2

Damage to the Facility
The main damage to the facility was identified as follows:(1) Collapse of Fuel Oil Storage Tank in Unit 1
The fuel oil storage tank that was intended to provide fuel for the auxiliary
boiler collapsed, and heavy oil leaked from the intake of Unit 1 towards the seaward
side. Adsorption mats were used to absorb the heavy oil and oil fences were deployed toprevent the diffusion of the heavy oil into the bay. When the tank collapsed, theauxiliary boiler was not operating so fuel oil was not being supplied.
The installation location of this tank was at O.P. +2.5 m. It collapsed as a resultof being struck by the tsunami.
(2) Fire in M/C of Unit 1
After the earthquake, a fire alarm was activated in the main control room. The
operators went to the site and found smoke being emitted from the basement of theturbine building at 15:30. The specific source of the smoke could not be confirmed due
to poor visibility. Later, however, it was determined that the M/C of the regular system
was burning. The communal fire service was not able to reach the site as the roads of
the Oshika Peninsula were impassable as a result of the earthquake and tsunami.
Therefore, the internal fire brigade tackled the fire, which was confirmed as having been
extinguished at 22:55 on March 11.
The breaker that tripped in the connection section of the M/C had been
violently shaken by the earthquake, as shown in Figure 6.2-3. The disconnect part ofthe breaker was damaged and came into contact with other parts in the M/C, causing
short circuit. The resulting arcing melted the insulation of the cables, generating thesmoke that was observed. Subsequently, a different breaker of a more earthquake-proof
design was installed.

117

Fig. 6.2-3 Onagawa Unit 1 Results of examination of cause of high-voltage power
supply board fire(3) Inundation of Reactor Auxiliary Cooling System Pump B and HPCS Auxiliary
Cooling Water System Pump of Unit 2
A plant walk down conducted after the earthquake confirmed that seawater hadentered the heat-exchanger room of the auxiliary cooling system in a non-radiation-controlled area of the third basement of the Unit 2 reactor building. The reactor
auxiliary cooling system pump (B) and HPCS auxiliary cooling system pump were bothsubmerged and failed as a result. Pump (A), however, was not affected and was used tocool the reactor and SFP.
The reason for this flooding was believed to be as follows: The sea surface rose
as a result of the tsunami. The lid of the water gauge box in the seawater pump room
was pushed up, allowing seawater to enter and flow through underground openings suchas those points where cable trays, piping and so on penetrate the walls. It was assumed
that water entered the heat-exchanger room of the reactor auxiliary cooling systemthrough the piping penetrations, as shown in Figure 6.2-4. To prevent this fromrecurring, the tide gauge box was removed, and a closure plate was installed.
118

Fig. 6.2-4 Onagawa Unit 2 Route of seawater flowing into auxiliary building ofreactor building

6.2.3
Successful Operation and Efficacy of Safety Functions of Onagawa(1) Height of the Site
When the design of Unit 1 was first discussed, tsunami countermeasures wererecognized as being an important issue from the beginning of the planning stage. The
Tohoku Electric Power Company organized an internal committee that included outside
experts in civil engineering and geophysics, who repeatedly discussed the issue. As aresult of literature studies and interviews with residents of the area, it was assumed thata tsunami near the site could reach a height of about 3 m. But discussions by thecommittees concluded that: (1) the height of the site would be the most effective meansof surviving any tsunami; (2) a height of O.P. +15 m was appropriate. Thus, it wasdetermined that the height of the first floor of the important civil structures and mainbuildings would be O.P. +15 m, and the site height would be O.P. +14.8 m. The
postulated tsunami height was reviewed several times after the construction of Unit 1,such as when the applications for the reactor establishment permits were made for Units
2 and 3, and upon the publication of a tsunami estimation method by the JSCE.
Subsequent re-evaluation did not produce a tsunami height that exceeded the height ofthe site, however. Even though the site height fell by 1 m (to O.P. +13.8 m) as a result of
movement of the earth's crust caused by the earthquake, the tsunami still did not exceed
the height of the main buildings. Therefore, the important safety functions of the
facilities were assured, allowing the reactor to enter cold shutdown status.
(2) Strengthening of Seawalls
As part of the application for the reactor establishment permit for Unit 2, a119

numerical simulation was undertaken to review the tsunami height, with the result thatthe postulated tsunami height was raised from O.P. +3 m to O.P. +9.1 m. To enhance thestability of the slope of the site against a tsunami undertow, the slope was covered withconcrete blocks up to a height of 9.7 m. The aim of this measure was to prevent damage
from subsequent waves after an initial tsunami had struck the site.
(3) Design of Seawater Pump RoomAt the Onagawa, instead of being installed in a harbour close to the sea, theseawater pumps were installed in pits, about 13 m deep relative to a site height of O.P.
+13.8 m. This structure will not be submerged provided a tsunami does not exceed thesite height. The tsunami that followed the earthquake had a height of O.P. +13 m andtherefore did not exceed the site height, and therefore the emergency seawater pumps
were not flooded. As described Clause 6.2.2, however, some pumps failed as a result ofwater ingress.
(4) Improvement of Seismic Safety Margin
To further improve the seismic safety margin, work to seismically strengthen
the facilities had been voluntarily undertaken at Onagawa. This work included thereinforcement of the supporting structures for the piping and electrical cabling. A totalof 6,600 locations in all three units had been addressed by June, 2010. This also
contributed to the relatively low level of damage in the facility.
6.2.4
Conclusion- After the earthquake struck, one of the five external power supply lines was
secured. All the EDGs were functioning normally.
- The tsunami, with a height of O.P. +13 m, did not reach the main buildings as
they had been constructed at a height of O.P. +13.8 m.
- In Unit 2, the reactor auxiliary cooling system pump (B) and HPCS auxiliary
cooling system pump were flooded and disabled, making EDG (B) and EDG for
HPCS unusable. However, the reactor auxiliary cooling system pump (A)continued to function.
- All the safety functions required to cool the reactor core and SFP performed well.
The safety of the plant was assured.
(Drafter: Tatsuyuki Sakuraba and Kosuke Kato)
120

6.3

Tokai No. 2 Nuclear Power Station
Tokai No. 2 Nuclear Power Station (Tokai Daini) of The Japan Atomic Power
Company (JAPC) is located in Tokai Village, Ibaraki Prefecture. Tokai Daini is a
BWR-5 with a MARK-II containment vessel. It has a generating capacity of 1100 MW.
Construction started in 1973 and the plant started commercial operation in 1978.
When the Tohoku-Pacific Ocean Earthquake struck, the main generator
automatically stopped in response to a turbine high vibration trip signal, after which thescram procedure was automatically implemented for the reactor. The seismic intensity
in Tokai Village was a 6 lower on the Japanese seven-stage seismic intensity scale.
Because of this seismic motion, the external power supply was lost and one of
the three seawater pumps for the EDG was disabled by the subsequent tsunami. The
other important safety systems continued to operate, however, and cold shutdown of thereactor was achieved at 00:40 on March 15, 2011, four days after the earthquake.
Detailed information on the Tokai Daini facility is given in reference [6-5], anda summary was given in the Seminar [6-6].
6.3.1

Summary of Earthquake and Tsunami
The maximum acceleration recorded in the basement of the reactor building
was 225 gal in the horizontal direction (east-west), and 189 gal in the vertical direction,
which was less than response acceleration of the design basis ground motion S S. The
response spectrum was somewhat lower than that based on Ss. Table 6.3-1 lists the
maximum accelerations recorded in the reactor building.
At 16:50 on March 11, the highest tsunami wave was recorded at H.P. +5.5 m.
As a result of an investigation of the damage left by the tsunami around the facility, andconsidering the diastrophism caused by the earthquake, the height of the tsunami wasestimated to be around H.P. +5.7 to 6.2 m, with the upstream portion of the wave
reaching H.P. +6.2 m. (Note: H. P. ±0 m is the datum level of Hitachi port,
corresponding to -0.89 m relative to the standard mean sea-level of Tokyo Bay.)
6.3.2
Impact of Earthquake and Tsunami(1) Impact of Earthquake on the Facility
Seismic damage was observed in some equipment such as the turbine system,
for which seismic resistance is less important, so called class B and class C. No damage
was found in those systems for which earthquake resistance is paramount, so calledclass S related with new Seismic Design Guide.
121

Table 6.3-1 Maximum acceleration based on acceleration wave data observed in
the reactor building
Location

6F

4F

2F

B2F

No.

RB04

RB03

RB02

RB01

North-South

492

301

225

214

East-West

481

361

306

225

Up-Down

358

259

212

189

North-South

932

612

559

520

East-West

951

612

559

520

North-South

799

658

544

393

East-West

789

672

546

400

Up-Down

575

528

478

456

Record(cm/s2)
Initial Design(cm/s2)
Design Basis (SS)*(cm/s2)* The maximum response acceleration at the each floor of the reactor building, based on
the basis seismic motion of 600 gal on the free rock surface at E.L. -370 m of theelevation level

JAPC assessed the soundness of those systems that must be earthquake
resistant, based on the seismic records for each floor of the reactor building. The
maximum acceleration recorded in the reactor building did not exceed either theacceleration of the initial design seismic wave assumed for the approval of construction
plan at the time of construction, or the maximum response acceleration spectrum ofdesign basis ground motion Ss.
The floor response spectrum measured on each floor partly exceeded the floor
response spectrum for the initial design seismic wave for the period between 0.65 s and
0.9 s. However, for the natural periods of the major equipment and piping, that areimportant to seismic safety, the measured floor response spectrum did not exceed theinitial design. Therefore, the deformation of that equipment and piping would haveremained in the elastic state.
(2) Impact of Tsunami on Facility
At 19:52 on March 11, five hours after the earthquake, EDG cooling seawater
pump 2C stopped automatically as a result of the north emergency seawater pump room
being flooded by the tsunami. Figure 6.3-1 illustrates the work that was done to seal the
points of penetration into the emergency seawater pump room. To protect the122emergency seawater pump room from tsunami, a new sidewall with a height of H. P.
+7.00 m was built outside the existing sidewall (H.P. + 5.80 m). Sealing work was
implemented between the two walls to prevent water ingress. At the north side,
waterproofing work for the above-ground cable pit lid was performed to prevent the lid
from being lifted by a tsunami. The arrangement here, however, made it easy for waterto build up. As a result, water flowed into the space between the two walls, and thenflowed over the previous sidewall and into the emergency seawater pump room.
However, given the redundant design concept, with EDG cooling seawater
pumps and RHR cooling seawater pumps provided separately to the north and south ofthe emergency seawater pump rooms, the EDG cooling seawater pump and the RHR
cooling seawater pump in the south room remained operational.
The operation of diesel generator cooling seawater pump 2C was restored 10days after it was submerged.
6.3.3

Situation at the Facility After the Earthquake
The earthquake caused the turbine axis to vibrate violently, triggering theautomatic shutdown of the main generator. Subsequently, the closure of the main steam
stop valves caused scram processing for the reactor to be implemented automatically. In
addition, the external AC power supply was lost. This was provided through two 275 kV
regular lines and a 154 kV backup line. All three of these lines were lost as a result of
the seismic motion. In response, the three EDGs in the facility automatically started tosupply power to the safety equipment needed to shut down the reactor.

Fig. 6.3-1 Implementation of sealing to points of penetration into emergency
seawater pump room123

Parameters related to the reactor core after the earthquake is illustrated in
Figure 6.3-2. The loss of external power meant that the condensate and feedwater
systems could not be used to remove the decay heat from the reactor core. Therefore,high-pressure steam was released via the SRVs to the suppression pool, and two RHRsystems were used to cool the water in the S/C to remove the decay heat.
Regarding the water revel in the RPV, the level in the reactor core immediately
fell as a result of the closure of the main steam stop valves and the loss of the feedwater,which caused both the HPCS and RCIC to start up automatically. The water level andpressure in the RPV were controlled using the RCIC and the SRVs, allowing the reactor
to be gradually depressurized.
At 19:25 on March 11, EDG cooling seawater pump 2C was stopped
automatically due to the tsunami. In response to this, the operators stopped a RHR (A)and an EDG (2C). The other, available EDG was able to maintain the power supply
needed for decay heat removal and so the depressurization of the reactor core could becontinued.

Fig. 6.3-2 Reactor parameters from occurrence of earthquake to cold shutdown124

On March 12, restoration of the external 154 kV external power supply line
was started, and power was successfully restored at 19:41 on March 13. After the
restoration of the RHR, cold shutdown was achieved at 00:40 on March 15.
The PCV parameters after the earthquake are shown in Figure 6.3-3.
Regarding the cooling of the PCV, due to the loss of external power, the EDGs weredrawn on to provide power for the safety equipment. The two RHR systems were used
to cool the suppression pool and control the pressure and temperature in the PCV. The
water level in the S/C increased immediately after the earthquake, as a result of coolant
being fed into the reactor via the HPCS and the RCIC from the CST. The water in S/C
was carried to the radioactive waste treatment system via the RHR pump, but the water
could not be treated due to the loss of the external power. Power to the radioactive waste
treatment system was also required to restore the water level in the CST. Although
low-voltage power supply cars were delivered after the earthquake and an air-cooleddiesel generator, provided for the decommissioning of the Tokai Power Station, wasavailable, none of these were actually used. Rather, the gas-turbine power generator on
the roof of the seismically isolated ERC building, which was installed in response to thelessons learned from the Chuetsu-oki earthquake, was used because connection to theequipment was easier. Then, water from the S/C could be treated and pumped back into
the CST.

Fig. 6.3-3 PCV parameters after earthquake
125

Although one EDG was disabled by the tsunami and one of the RHR systems
became unavailable, safety was maintained due to the redundancy of the safety systems.
It would have been possible to achieve cold shutdown quickly through the restoration of
the AC power to the emergency systems by sharing the power from the functioning
diesel generator and then starting up the RHR system. There was a risk, however, of theremaining diesel generator failing to start, and the redundancy of the high-pressure
injection systems being lost. In fact, in shutdown cooling mode, the RHR system wasstarted up as soon as the external power supply was restored, rather than incurring therisks associated with an earlier cold shutdown, because the parameters of the plant werestable and safe, and the restoration of the external power was expected within a few
days. Cold shutdown was achieved at 00:40 on March 15.
While determining the procedure to be applied to achieve cold shutdown,
damage to the seawater pumps was investigated even though the major tsunami warning
had not yet been lifted. This investigation confirmed flooding in the north side but thatthe south-side seawater pump room was secure. The staffs drained the seawater from theflooded room to prevent the extension of the damage. After discussing the risks
associated with implementing a rapid cold shutdown of the reactor, the procedure wasdetermined.
6.3.4

Countermeasures against Postulated Tsunami
Since the application for the reactor establishment permit was made for Tokai
Daini, the evaluations of and countermeasures against tsunami have been improved
many times. When the application for the establishment was first made, there was nostandard related to countermeasures against tsunami. Based on the literature available atthe time, the highest tide in the vicinity was recorded at H.P. +2.35 m. This was a resultof the 1958 Kanogawa typhoon, and was taken as the design basis for any possible
tsunami at that time.
The tsunami that struck after the 1993 Hokkaido Nansei-Oki earthquake led to
the Government starting discussions related to the issue of guidelines for responding totsunami. JAPC also carried out an internal review of how to respond to tsunami and, asa result, raised the postulated tsunami height to H.P. +5.3 m. Then, JAPC built a
protective wall of H.P. +5.8 m in height around the north emergency seawater pump
room. When JSCE published its tsunami evaluation method in 2002, JAPC applied thismethod and arrived at a maximum tsunami height of H. P. +5.75 m, which indicated thatthe implemented prevention measures would be able to resist a tsunami.
In 2007, Ibaraki Prefecture identified those areas that it assumed would be126flooded by a tsunami. JAPC used this information as new knowledge to evaluate andperform a back-check of its seismic safety. Based on the Ibaraki Prefectural evaluation,
the postulated tsunami height was estimated to be H. P. +6.61 m and the construction ofa side wall with a height of H.P. +7.00 m around the emergency seawater pumps wasundertaken. This was a major factor in protecting the important safety systems from thetsunami resulting from the 2011 Great East Japan Earthquake.
6.3.5
Conclusion(1) Immediately after the earthquake struck, all the external power supplies were lost.
However, scram processing for the reactor was successful, and the three EDGs werestarted to supply power to the safety equipment.
(2) The earthquake was followed by a tsunami with a height of H.P. +6.2 m, which wasmuch higher than the tsunami height that was postulated at the time of construction,
which was H.P. +2.35 m. However, continuous improvement of the tsunami
countermeasures, including a review of the tsunami estimates to reflect new
knowledge, and the modification of the facilities, resulted in the important safetyfunctions being maintained after the tsunami struck. Two of the three EDGs
continued to operate through the crisis, although diesel generator (2C) failed as aresult of flooding caused by the tsunami.
(3) The water injecting function was maintained and the function of one RHR was
secured. As a result, cooling water could be injected into the reactor core, anddepressurization of the reactor core could be performed steadily, with decay heatbeing removed until cold shutdown was finally achieved.
(Drafter: Noriaki Tomura)

127

6.4
6.4.1

Comparison of Events in Nuclear Power Plants
Viewpoint for Comparative Analysis
The earthquake and the tsunami occurred over a wide area including four
nuclear power stations. The events were managed and settled safely at Onagawa,
Fukushima Daini, and Tokai Daini. Further, the Fukushima Daiichi Units 5 and 6 also
managed to avoid severe accidents, but Units 1 to 4 could not.
When an abnormal transient and/or accident occur, cooling and heat removing
functions were necessary to be sustained to avoid severe accident. All plants who could
settle the situation after the earthquake could secure these functions somehow. We alsoneed to focus on the surviving plants to obtain knowledge and lessons.
In this section, preparedness against the earthquake and tsunami aresummarized at first, and the lessons learned from the comparison are investigated.
Serious accidents are caused by the superposition of unrecognized faults andadventitious events. Latent problems are often recognized after accidents occur. Lack ofknowledge and/or experiences is a major cause in this case. If a recognized doubt was
ignored, this is the serious situation lacking in safety culture and cannot be allowed.
Mistakes in safety design, safety management, and safety regulation have beenfound by investigations of the Fukushima Daiichi accidents. Two major mistakes are as
follows:
1) Lack of countermeasure against external initiators, especially tsunamis. This caused
station blackout, loss of ultimate heat sink, and loss of DC power. The earthquake
and tsunami also disturbed the restoration activities after the accident occurred.
2) Several accident management measures have been prepared to prevent accidents
from growing more severe and/or to mitigate the effect of severe accidents.
However, some of these measures are difficult to operate under real accident
conditions.
These findings are easily obtained from the investigation of the Fukushima
Daiichi only. In this chapter, however, damage to the equipment in other plants are also
investigated and categorized to draw notable conclusions. We studied how theearthquake and the tsunami destroyed the equipment, and how the advanced safetymeasures mitigated their effects. The following assumptions were considered:
1) What if the tsunami wave striking the facility were higher? This could
have happened because the height accords to the superposing of waves
from different wave sources. In reality, the superposing was stronger at
Fukushima Daiichi than at Fukushima Daini.
1282) What if the tsunami had struck the facility earlier? Such large tsunamis
occur once in a thousand years or longer. It could have happened 20 years
ago, when neither corresponding actions against tsunami evaluation by
JSCE nor accident management measures existed.
3) What if the operation statuses such as in service, shutdown period, weredifferent?
In general, discussion based on such assumptions, the so-called ‘what if study’,should be concluded after appropriate analysis and evaluation. This report only proposes
the research subject. However, we hope that the further studies are conducted by nuclear
plant operators and research facilities for finding the essentials that lead to continuous
improvement.
We should develop reasonable and available countermeasures to mitigate theproblems in safety design, safety management, and safety regulation revealed by the
Fukushima Daiichi accident. For initiators with large potential hazard, suitablecountermeasures should be designed considering their intrinsic character. Setting designbasis hazards is an important issue for these activities. ‘What if studies’ will help us
decide appropriate design basis hazards.
6.4.2

Seismic Motion Observed in Each Power Station and its Effects
The seismic motion records observed in each nuclear power station and the
maximum response acceleration corresponding to design basis ground motion SS aregiven in Table 6.4-1. The damage to facilities caused by the earthquake is described in
Chapter 5 and Section 6.1~6.3. In this section, the main aspects regarding the damage
to facilities that is confirmed and related assumptions are described.
According to the seismic records of Fukushima Daiichi and Onagawa, somevalues of maximum acceleration recorded are greater than the maximum response
acceleration with respect to SS, but up to now, the safety function of the facilities thatare important in seismic design have not been confirmed. Some facility with lessimportance, so called class B and class C, has been confirmed to be damaged.
(1) Fukushima Daiichi
For the seismic motion at the Fukushima Daiichi, the maximum acceleration of
the east-west direction in Units 1, 2, and 5 is up to 25% greater than the maximum
response acceleration with respect to the design basis ground motion SS, while othervalues observed are lower than maximum response acceleration.
Regarding the influence of seismic motion, the transmission tower collapsed
129and the receiving breaker was damaged, which resulted in the loss of all external power.
With regard to the influence on reactor facilities, it is impossible to access theinteriors of Units 1~4; a walk down Unit 5 revealed no damage, except the damage
caused to one drain pipe around the moisture separator in the turbine building. The EDG
automatically started after the loss of external power. In addition, in Units 1~4, there
was no trend indicating damage in the reactor pressure boundary according to plantparameters. Estimations based on the reproduction analysis of seismic response showedthat the major safety equipment did not incur damage.
(2) Fukushima Daini
With regard to the seismic motion observed at Fukushima Daini, almost all thevalues were 40% to 60% of the maximum response acceleration with respect to SS. The
seismic motion was less than the maximum response acceleration at every location. The
maximum acceleration in the horizontal direction was 277 gal in Unit 3, and the
maximum acceleration in the vertical direction was 305 gal in Unit 1.
Two of the four external power supply lines were lost by the damage caused by
the disconnecting switch and/or lightning arrester. One of the four lines was under
maintenance. One line was secured, which helped the recovery activities considerably.
Furthermore, with regard to the impact of reactor facilities, on the basis of the
walk-down results, no abnormality in appearance was found in facilities important tosafety, while some equipment of seismic class B and C was damaged. In addition,according to seismic analysis, loads were below SS, and according to a piping systemanalysis, some results of the floor response spectrum exceeded SS but were lower thanthe evaluation reference value.
(3) Onagawa
Onagawa is the closest nuclear power plant to the epicenter, and is located at a
distance of 130 km from the epicenter of earthquake. The maximum accelerations
observed in Units 1~3 were slightly larger than the maximum response acceleration
with respect to SS, but it can be considered as almost equal to the response of S S.
In terms of external power, there were 5 lines in total: Matsushima Line (2lines), Oshika Line (2 lines), and Tsukahama Line (1 line). After the earthquake, oneline of the Matsushima Line was normal and supplied power to three units. The cause ofloss of external power supply was a ground fault of the arrester and damage to a ceramic
insulator connecting wire.
Concerning the impacts on nuclear facilities, fire broke out in the power panel
130

of Unit 1 because the start-up transformer was damaged. External power supply became
impossible despite the fact that one Matsushima Line was still working. The shortcircuit of the breaker caused the arc discharge in the panel, and ignited to cable insulator,and led the fire.
Other equipment important to safety was functioning normally until thetsunami. Impacts of the earthquake include dropping of a dry canister in a solid waste
storage place, falling of the ceiling in the main control room of Unit 1, foreign materials
falling into the SFP of Unit 3. None of these impacts was severe.
At the Onagawa, a large earthquake that exceeded the previous S 1 value had
occurred previously in 2005. As a result, the plant was shut down for a long period time
and the seismic margin was improved. These activities contributed to reducing seismic
damage.
(4) Tokai Daini
Based on the seismic observation records at Tokai Daini, most values are 40%
to 60% of the maximum response acceleration with respect to the Ss. Seismic records at
all measurement points were not excess the maximum response acceleration of S S.
Struck by the earthquake, all three external power lines were lost. However,one 154 kV transmission line was recovered at 19:41 on March 13. The reactor achieved
cold shutdown at 0:40 on March 15. On the other hand, the restoration of 275 kVtransmission line needed more time because the breaker at the substation was damaged.
Furthermore, towards the impact on reactor facilities, no damage was found in
equipment important to safety, while some equipment of seismic class B and C wasdamaged. EDG automatically started functioning after loss of external power.
In addition, many damages were found on heavy components; for example, ashielding door fell down, and damage was found in fastener of the shielding. The
seismic evaluation against some of the lower class equipment was only based on theseismic load. Further, connection points of the equipment were considered to bedamaged.
(5) Summary of Seismic Motion and its Impacts
A summary of the status of the four nuclear power stations is as follows.
In parts of the Fukushima Daiichi and Onagawa, the seismic acceleration
exceeded the maximum response acceleration with respect to their SS, while most wereless than the maximum response acceleration of SS. Therefore, design assumption was
not sufficient, which is one of the lessons learned. However, the earthquake impacts on131

the facilities beyond their seismic design were not serious. The tolerance in their design
contributed significantly to the safety. The way to set the safety margins beyond designbasis should be reconsidered when we discuss the countermeasures beyond the design
postulation.
In addition, major damage to the external power supply system was caused in
all the nuclear power stations by the earthquake. All seven power lines in Fukushima
Daiichi, two of the four lines in Fukushima Daini (one line was under maintenance),four of the five lines in Onagawa, and all three lines in Tokai Daini were lost. If external
power supplies were alive, many recovery activities and SA prevention measures could
have been implemented. However, since all external power lost, more effort wasinvested on recovery work. As described in Table 6.4-3, at Fukushima Daini (Clause
6.1.2), Onagawa (Clause 6.2.3), and Tokai Daini (Clause 6.3.5), the secure of external
power and EDG power made the recovery work easier to be performed. The reliabilityof the power system, including external power supply was desirable for the safe
shutdown of the reactor.
6.4.3

Tsunami Wave Observed in Each Power Station and its Effects
The postulated tsunami height on the application for reactor establishment
permit and its revisions at each nuclear power station are described in Table 6.4-2.
Details of the damage caused by the tsunami have already been explained in Chapter 5
and Sections 6.1–3. Here, the related facts and estimation of the damage caused by thetsunami are summarized.
All nuclear power stations have updated their postulated tsunami heights toimprove their tsunami countermeasures on the basis of newly acquired knowledge
and/or experience, as shown in Table 6.4-2. Utilities decide the postulated tsunami
height when they submit the application for reactor establishment permit. Revisions to
the postulated tsunami height were made in 2002 after JSCE issued a method forassessing the tsunami height. In Fukushima Daiichi, Fukushima Daini, and Tokai Daini,the utilities re-reviewed their tsunami measures in 2007 and adopted corresponding
activities in accordance with prefectural tsunami assessments for natural disasterprevention activities performed by the local government of Fukushima and Ibaraki.
The maximum height of the tsunami wave in each nuclear power station
assumed by the traces is described, along with the elevation of the main buildings, as
follows. In Units 1~4 of Fukushima Daiichi, the maximum wave height was about O.P.
+13.1 m, and the elevation of the main building area was about 10.0 m, where O.P.
implies the standard level at Onahama Point in Fukushima Prefecture. In Fukushima132

Daiichi Unit 5~6, the maximum wave height was O.P. +9 m, and the elevation of themain building area was about 13 m. In Fukushima Daini, the maximum wave height
was O.P. +7~8 m, and the elevation of the main building area was about 12 m. In
Onagawa, the maximum wave height was about O.P. +13.0 m, where O.P. implies thestandard level at Onagawa Port. The elevation of the main building area in Onagawa
used to be O.P. +14.8 m before the earthquake, but it decreased up to O.P. +13.8 m. In
Tokai Daini, the maximum wave height was H.P. +5.7~6.2 m, and the elevation of themain building area was about H.P. +8.9 m, where H.P. indicates the standard level at
Hitachi Port in Ibaraki Prefecture. The maximum wave height did not exceed theelevation of the main building area, except in Fukushima Daiichi Units 1~4, where coredamage and/or a significant accident occurred.
The flood height around all the main buildings was less than 2.5 m in
Fukushima Daini and was almost zero except in Unit 1. There was no flooding of themain building areas in Onagawa and Tokai Daini.
(1) Fukushima Daiichi and Fukushima Daini
Tsunami postulation in Fukushima Daiichi and Fukushima Daini was first
decided considering the tsunami caused by the Chili Earthquake in 1960. Applying theassessment method proposed by JSCE in 2002, the postulated tsunami height wasincreased to 5.7 m in Fukushima Daiichi and 5.2 m in Fukushima Daini. Seawater
pumps and buildings below these elevations were improved by lifting up to a sufficientheight and/or tightening the waterproof seal. The tsunami height in Fukushima
Prefecture was originally assessed in 2007 to enhance the efficiency of countermeasures
to natural hazards. No additional countermeasure was adopted at that time because thetsunami height in these nuclear power stations, assessed by Fukushima Prefecture, waslower than the postulated tsunami height evaluated by JSCE’s method. Thereafter,
coordination of the ocean bed and tide conditions was reviewed, so that the postulated
tsunami height was increased to 6.1 m in Fukushima Daiichi. This led to additional
lifting of the seawater pumps.
Newly acquired knowledge on natural disasters was not just used to revise thepostulated tsunami height. For example, TEPCO considered the opinion by the HERP in
2002. Accordingly, the Pacific Ocean Plate may cause a huge earthquake off the coast of
Fukushima, a possibility that had not been considered in previous studies. As a result of
trial calculations reflecting this possibility and other new knowledge, the maximum
tsunami height was estimated to be 8~9 m. Similar to the hazard in Fukushima Daini, atsunami of this level can run-up towards the main building area. As expected by HERP,
133

motion of the Pacific Ocean Plate caused a huge earthquake that led to the Fukushima
accident. However, HERP did not consider the synchronized motion between manysources. Further, the trial calculation by TEPCO did not consider this possibility.
Since the tide gauges were broken by the tsunami, the reported tsunami heightat the Fukushima Daiichi and Fukushima Daini is based on estimation. As a result of theestimation using the wave source model, which is reported in Section 5.1, the maximum
height at the Fukushima Daiichi and Fukushima Daini is about O.P. +13 m and O.P.
+7~8 m, respectively. Superposing of waves from multiple wave sources caused thesedifferences. The tsunami was not caused by a single wave source model. It was causedby the combined motion of multiple wave sources modelled in JSCE’s method,
including the wave source model of Jogan earthquake. In addition, the tsunami moved
upstream at the site and reached up to O.P. +15.5 m and O.P. +14.5 m of each nuclear
power station.
These fact shows that we should consider that the tsunami can go upstream in
the site, in addition to the synchronized motion of multiple wave sources, superposingof waves from each source, and postulated tsunami height.
Following are the damage caused by the tsunami in the Fukushima Daiichi and
Fukushima Daini.
The damage is more severe in Units 1~4, where the elevation around the mainbuildings is 10 m, as opposed to that in Units 5~6, where the elevation is 13 m. All the
EDGs in Fukushima Daiichi were lost, except for system B in Unit 6. Automatic
starting of the EDGs was successful after loss of external power. Though the EDGs ofsystem B in Units 2, 4, 6 were located on the ground floor, and the rest were located in
the basement. Though the EDGs of system B in Units 2 were located on the groundfloor, and the rest were located in the basement, indirect factors such as loss of support
cooling systems and/or submerging of the M/C caused loss of function of the EDGs, inaddition to submerging of the EDGs themselves,
In Fukushima Daiichi, a single EDG was shared between two units before the1990s. To avoid this sharing, additional EDGs were equipped in Units 2, 4, and 6. These
additional EDGs were of the air-cooled type because it was difficult to design sea-water
piping. These EDGs were located at a greater height than the original ones and hence
were not submerged. Further, the EDG in Unit 6 survived the tsunami because therelated equipment, too, was not submerged. When DC power is lost or fully discharged,
it becomes difficult to control DC-driven valves, comprehend important parameters, andcontinue the operation of pumps. DC power was lost by flooding inside the buildings in134

Units 1, 2, and 4; further, DC power was fully discharged in Unit 3 but remained
uninterrupted in Units 5~6. All the seawater systems were lost in all six units.
In Fukushima Daini, flooding occurred in many buildings because of leakage,by the destruction of the equipment entrance, and through the concrete ducts. In the heatexchange building for system B in Unit 3, the equipment entrance was not destroyed by
the tsunami, and the equipment inside remained secure.
As a result, all the EDGs in Units 1 and 2 were lost. One of the three EDGs
was lost in Unit 3. Two of the three EDGs were lost in Unit 4. DC power was secured inall units. All seawater systems in Units 1 and 2 were lost. In Unit 3, the seawater system
for the EDG of the HPCS was secured, but that for half of the RHR system was lost. In
Unit 4, the seawater system for the EDG of the HPCS was secured, but that for all of the
RHRs was lost.
By considering whether the following facts are fortuity or necessaryconsequence, we can obtain useful information to improve safety design.
Fukushima Daiichi
・ Equipment function was lost because of the external initiator, tsunami.
Fukushima Daini・ In Units 2 and 4, the EECW pump for system A was located on the first floor, whilethat for system B was located on the second floor. Only system B survived thetsunami.
・ In both Unit 2 and Unit 4, two RHRS pumps for System B were equipped closely.
One of the pumps became non-functional in both units.
・ The power distribution board in the reactor buildings was secured, except in Unit 1,
where flooding occurred.
・ The EDG for system B and the HPCS in Unit 3, and the EDG for the HPCS in Unit
4, which were not submerged, were unaffected.
(2) Onagawa
When the application for reactor establishment permit was submitted in 1970,the postulated tsunami height was about 3 m. Through a discussion that took experts’opinions into account at the internal committee of the Tohoku Electric Power Company,
the vice president at the time strongly advocated setting the elevation of the site at O.P.
+14.8 m. Countermeasures against the backwash of tsunamis were prepared from the
beginning of the establishment. For example, the seawater pump room was designed to135secure a water source for a certain time even if a backwash occurs. In addition, in lightof new knowledge, extra measures such as the strengthening of the slope of the area
were taken.
Eventually, the Tohoku-Pacific Ocean Earthquake caused the site to subside.
After the earthquake and the accompanying tsunami of height 13 m, the elevation
became 13.8 m.
The tsunami’s effects, such as the collapse of a heavy oil tank in Unit 1 and theflooding of seawater into the annex of the reactor building in Unit 2, were limited.
Comprehensive protective measures against tsunamis were quite effective in Onagawa.
Even though the designs of Units 2 and 3 are almost identical and theirelevations were sufficient against the tsunami, flooding occurred only in Unit 2. The
seawater flowed into the building through the tide gauge connection at the seawater
pump room. The objective of the tide gauge is to trip the turbine when the tide reaches acertain level. The gauge in Unit 3 was originally prepared when the reactor
establishment application was submitted. Following this design, a similar gauge wassubsequently equipped in Unit 2.
Though the gauge belongs to a regular system, the gauge in Unit 2 was locatedin the area for the emergency seawater system because of space constraints. Anaseismof the tsunami pushed up the storage box, and the auxiliary cooling system heatexchanger room was flooded through the opening for the gauge. Consequently, thepumps for nuclear reactor auxiliary cooling systems (System B and System H) were lost.
Hence, EDG for system B and EDG for system H were lost. Consideration of theanaseism pressure was insufficient when the gauge in Unit 2 was added. Even though
the elevation of the site was higher than the highest tsunami wave, some emergency
systems were damaged by the modification of regular system.
In the case of Fukushima Daiichi Unit 6, an additional EDG was equipped at
higher position than the others because of space constraints, and it survived from thetsunami. In contrast, in the case of Onagawa Unit 2, an additional tide gauge was
equipped on the emergency system area because of space constraints, and it led to theflooding of the area. Though both activities aimed to improve the safety of the plant, theresults were contradictory. This contrast shows that the location and coordination ofequipment are essential especially when installing additional equipment and back-fitting
the plant design. An important point inferred from the Tohoku-Pacific Ocean
Earthquake is that the location and coordination of equipment should be considered
carefully.
136

(3) Tokai Daini
The postulated tsunami height in Tokai Daini was H.P. + 2.35 m when thereactor establishment permit application was submitted. Countermeasures for tsunamis
had been strengthened in light of the JSCE method in 2002 and the tsunami assessment
for natural disaster prevention by the Ibaraki Prefecture in 2007. The tsunami height in
the Tokai Daini in the Ibaraki Prefecture’s assessment was higher than the height
evaluated using JSCE method. On examining the assessment, JAEC, the plant utility,
recognized the result as new knowledge for back-checking the seismic measures.
Because of the tsunami of the Tohoku-Pacific Ocean Earthquake, the seawater
pump room at the northern side was flooded. Consequently, a seawater pump for EDG
stopped automatically and EDG 2-C was stopped manually. One of the RHRs was also
lost. In contrast, the seawater pump room at the southern side was not flooded. As aresult, one of the three EDGs was lost even though all EDGs successfully started
operation when the external power supply was lost. The conditions in the northern andsouthern areas differed because construction work for raising the wall of the pump room
was underway when the tsunami struck; this is reflected in the assessment by Ibaraki
Prefecture. The construction in the southern area was completed, but it had not finishedin the northern area. A hole penetrating through the wall had not sealed yet, and thewater entered through the hole. A huge accident at Tokai Daini was avoided because
continuous improvement measures were taken on time.
(4) Conclusion of the Tsunami and its Effects
In the tsunami assessment, waves from different wave sources had not beensuperposed. Such superposition must be considered in the assessment of tsunami
countermeasure in accordance with the discussion in NRA.
Flooding occurred even in plants located at elevations higher than the
maximum tsunami height. The tide gauge in Unit 2 of the Onagawa and the penetrationof the wall in the Tokai Daini were the routes for the flooding. The importance of tidewalls and water tightening doors had already been widely recognized. However, watercan enter through the openings. In light of the tsunami incidents, we should
continuously investigate measures to perfectly protect facilities from flooding andminimize the effect of flooding.

1376.4.4Site Conditions after the Earthquake and Tsunami and the Recovery
Operation
The damage in the facilities was explained with their design information in
Clauses 6.4.2 and 6.4.3. In these clauses, the situation in the facilities, responses ofoperators, and recovery actions are described. The situation of the equipment in thefacilities is summarized in Table 6.4-3.
(1) Fukushima Daiichi
In Fukushima Daiichi, Units 1~3 were in operation, and Units 4~6 were in
regulatory maintenance. The hazard was summarized as follows: External power was
lost in all six units. An air-cooled EDG survived in Unit 6. Other EDGs were lost
because of the tsunami. All DC power sources were lost because of the tsunami in Units
1 and 2.
Because of these damages, the facility was left in the following situation.
The investigation of Unit 1 showed that the loss of DC power was a little fasterthan the loss of AC power. Owing to the loss of the pipe rapture detector, the interlockof the IC functioned, and PCV isolation valves on the pipes connected with the IC wereclosed using the residual AC power. This led to the loss of IC function, which can
reduce reactor pressure and sustain the reactor water level. Loss of DC power had beenrarely assumed when AC power remained. Operators may have thought that theisolation valves on the IC were maintained when the power had been lost; hence, theymight not have realized that the valves had been closed. Consequently, core damage
occurred early. Hydrogen, generated inside the damaged core, leaked from PCV to thereactor building. This caused the hydrogen explosion in the reactor building.
In the case of Unit 2, operators started/stopped operating RCIC manually tocontrol the reactor water level. Just before the tsunami struck, an operator re-started
operating the RCIC manually. Unit 2 lost DC power like Unit 1, and the plantparameters were lost. Though the DC power was lost, isolation valves of the RCIC kept
opening, and the RCIC had unexpectedly been driven by a two-phase flow to injectwater for about three days. During this grace period, operators tried to use AM measures
to inject water but did not succeed. Since the unstable operation of the SRV was affectedby the hydrogen explosion in Unit 3, water injection using fire engine was insufficient.
PCV venting failed, too. Finally, the reactor core in Unit 2 was also damaged. If an
operator had not started operating RCIC just before the tsunami struck, reactor coredamage in Unit 2 could have occurred faster. Hydrogen explosion did not occur in Unit
2. The reason for this is estimated to be that hydrogen was not contained in the reactor
138building because the blowout panels on the reactor building were opened by thevibration of the hydrogen explosion in Unit 1.
In the case of Unit 3, core cooling was maintained for 2 days using RCIC and
HPCI because the DC power was functioning. However, the DC power was fullydischarged by the loss of AC power. Finally, core damage in Unit 3 occurred faster thanin Unit 2. Hydrogen generated in the damaged core leaked from the PCV, which led tohydrogen explosion in the reactor building.
When the earthquake occurred, Unit 4 was in the regular maintenance period
and all nuclear fuels had been removed from the reactor. Therefore, the case of Unit 4 is
not a nuclear reactor accident. The decay heat of the spent fuels in the pool was large
because all fuels were inside the pool. Loss of the cooling function and of the water
feeding function for the SFP was worrisome. In fact, exposure of the fuel did not occur,as the water in the SFP was estimated to remain until the end of March. Hydrogen
explosion occurred also in Unit 4. It was suspected that the spent fuels melted in the
SFP the first time, but a later investigation showed that the hydrogen generated in Unit 3
flowed into Unit 4.
Unit 5 and Unit 6 were under cold shutdown. Unit 5 was in the last period ofcold shutdown, and refuelling had already been done. Unit 6 was under cold shutdown
for a long time due to some troubles. Therefore, decay heats in those units were small.
These units had a relatively longer time to settle the situation than Units 1~3. Though all
EDGs were lost in Unit 5, AC power was supplied to Unit 5 from an air-cooled EDG
surviving in Unit 6 through the electricity sharing line between Unit 5 and 6, which was
installed as part of accident management. DC power was secured in both Units 5 and 6.
Water injection using MUWC, which was also a part of accident management, anddepressurizing using SRV were performed on March 13.
Seawater systems were lost in all units. Later, in Units 5 and 6, the seawater
system was restored to recover the SFP cooling function. A power generation car was
utilized to power the provisional submersible pumps. Seawater systems were secured
from the seismic motion but were affected by the tsunami. All seawater pumps inFukushima Daiichi were lost. Since all AC power was lost, utilization of power
generation cars was tried. Though it was obstructed by the hydrogen explosion in theneighbouring plant at Unit 2, and by submerging of power distribution panels in Unit
1~4, the power generation cars were sometimes used effectively.
(2) Fukushima DainiIn the Fukushima Daini, all four units were under operation at rated power. One139

of the four external power lines survived and no station blackout occurred at this site.
All EDGs were lost in Units 1 and 2, but two of the three EDGs survived in Unit 3 andone of the two EDGs survived in Unit 4. All DC power was secured.
In the case of Unit 4, for example, since the seawater system lost function after
it was struck by the tsunami, reactor cooling using RCIC and depressurizing using SRV
were performed. At 18:33 on March 11, a special report according to the Article 10 of
Nuclear Emergency Act was issued because the functioning of the emergency apparatus
cooling system pumps was not recognized. Next morning, the temperature of the
suppression pool exceeded 100 °C, the event referred in Article 15 of Nuclear
Emergency Act. After seawater pumps were recovered and the emergency apparatus
cooling system and RHR started functioning, the restrictions referred in Article 10 and
15 of Nuclear Emergency Act were lifted.
To restore electric power, power was supplied from the waste treatmentbuilding and heat exchanger building of Unit 3, where the power distribution board was
not submerged, using power generation cars. Provisional cables, with lengths of about 9km, were extended by 200 people in a day.
(3) Onagawa
In Onagawa, Unit 1 and Unit 3 were under operation with rated output, andUnit 2 started operation under a regulatory maintenance period. One of the five lines ofexternal power survived and the site did not experience a station blackout. All EDGs in
Unit 1 and Unit 3 were secured. Two of the three EDGs in Unit 2 were lost due to theloss of the reactor auxiliary cooling system. All DC power was secured.
In the case of Unit 1, for example, at 14:55 on March 11, a start-up transformer
was stopped by the ground fault in the M/C of system A, which was caused by seismic
motion. Since external power for the regular system was down, core injection using
RCIC and depressurizing using SRV were performed. The cause of the failure of thestart-up transformer was the functioning of overcurrent relay affected by the ground
fault in the regular system A. After the inspection, at 2:05 on March 12, the start-up
transformer was restored, and regular systems without system A were recovered. While
the start-up transformer was non-operational, AC power was supplied from EDG toemergency lines, and the power source for emergency cooling systems of the reactor
and SFP was secured. The reactor was cooled using RHR, and was cold-shutdown at
0:58 on March 12.
Matsushima No. 2 line was only survived after the earthquake. Oshika No. 1line was restored at 20:12 on March 12, Oshika No. 2 line was restored at 20:15 on
140March 12, Matsushhima No. 1 line was restored on 10:47 on March 17, and Tsukahama
sub-line was restored at 15:41 on March 26, respectively.
(4) Tokai Daini
Tokai Daini was under operation with rated output. All of the three lines ofexternal power were lost by the earthquake motion. One of the three EDG was lost due
to the tsunami. Other safety functions were secured. The emergency battery for the
RCIC was charged using other electric power. DC power was secured.
No damage was observed in SSCs whose seismic importance was high. At
19:52 on March 11, seawater pump 2C for EDG cooling stopped automatically because
the northern emergency seawater pump room was flooded. It took about 10 days torecover operation. As a result, EDG-2C was stopped manually, and this led to the
shutdown of one of the RHRs. However, safety function was secured.
Regarding the cooling of PCV, external power was lost soon after theearthquake, but the power was supplied to safety equipment via EDG. Cooling of the
suppression pool using two RHRs was secured, and the pressure and temperature of
PCV were controlled. Core injection was performed using HPCS and RCIC pumps. The
initial water source of these pumps was a CST so that the water level in the S/C was
increased. The water treatment function of suppression pool was lost by the loss ofexternal power. Therefore, a gas turbine generator in the ERC building was used. This
seismically isolated building was constructed as per the experience of the Chuetsu-oki
Earthquake. The power was shared with the regular system to operate the waste
treatment system. Residual water in S/C was collected and treated by the waste
treatment system. In addition, a low-voltage power generator was coordinated to the site.
Air-cooled EDG for the Tokai Nuclear Power Station under decommissioning was alsoavailable at that time.
As described above, one of the three RHR systems failed after being struck by
the tsunami, but the safety function was maintained and the parameters of the nuclear
power plant remained stable. A 154 kV line for external power was restored on March
13, and a 275 kV line was restored on March 17. At 0:40 on March 15, cold-shutdown
was achieved using the cooling mode operation of RHR.
(5) Conclusion
Elevation of the site, design postulation of the SSCs important for safety, andpreparedness against the events beyond design basis were important factors. Securing
electric power source was essential to the progress of events and the management after
141

the severe accidents. The ultimate heat sink and preparedness of AM measures werealso important.
Loss of electric power was caused by various reasons. It seems impossible tofind a single bottleneck that reduced the reliability of the electric system. A
comprehensive approach should be adopted to improve the electric system. In addition,
to strengthen emergency electric systems against severe external initiators and toprepare external electric sources, portable electric power should be prepared.
As described above, seismic motion rarely damaged the SSCs important forsafety except for external power. This fact shows that the seismic design was almost
appropriate.
However, seismic motion led to the fire the power distribution board caught inOnagawa Unit 1. We should emphasize that the external initiator, the so-called seismic
motion, can lead to an internal initiator, i.e., the so-called fire. Consider the risk of fireand flooding, we should pay attention to these initiated by seismic motion and tsunami
in addition to random events.
It is the most important to avoid an event beyond design basis to achieve severe
accident. In case a severe accident occurs, it is important to utilize AM measures‘effectively’. This topic is discussed in Section 7.2.
6.4.5

Conclusion of the Comparison between Facilities
We discussed the effect of the Tohoku-Pacific Ocean Earthquake and theaccompanying tsunami on four nuclear power stations. Following is the conclusion of
the comparison.
At first, as described in Clause 6.4.1, it is essential to consider external
initiators to secure safety. It is difficult to find the appropriate postulation for protective
design against various external initiators. The problem on design basis hazard will be
discussed in Section 7.3. It is important to conduct the deepest consideration formeasuring protective design, and apply the result completely even though it seems atiny measure. For example in Onagawa, various measures had already been adopted.
Handrails were prepared in front of the controlling boards to observe the parameters
while the plant was shaking. Pit-type seawater pump room was adopted to secure water
source even if a large back rush occurred. In the sixth Seminar, an audience fromoutside commented that ‘This is my private opinion, but I though Onagawa could avoid
large damage inevitably’.
Though maximum effort was invested, human insights are limited. Consideringthat appropriate safety design is adopted taking into account all possibilities, we should142prepare flexible measures ‘considering the occurrence of inconsiderable events’. The
preparedness of portable power source is a good example of these measures. This issue
is considered in Section 7.2.
We would also like to emphasize that ‘safety principles that had been thought
to be important are still important’.
The first principle is DiD. This will be discussed in Section 7.1.
The most important lesson learned from the comparison between facilities may
be the importance of “continuous improvement”. Regarding this point, we would like tointroduce the comment from the audience from outside NSD at the sixth Seminar.
“Listening to the presentations, I thought that continuous activities, such as theconstruction in Tokai Daini based on the tsunami assessment and backrush preparedness
in Onagawa were effective. It must be difficult for both facilities to perform these works,
but they did. Even though there were some lacks of consideration or incomplete
fabrications, I think that these activities led to settle the situation”.
Bad experiences are easily remembered, but good experiences are often
forgotten. In the case of Fukushima Daiichi Unit 5, the plant did not suffer because thepower line shared with Unit 6 had been prepared. If the earthquake and tsunami had
struck the site 20 years ago, the situation at Fukushima Daiichi Unit 5 and Tokai Dainiwould have been worse than now. This is an example that continuous improvement
occurred.
We would like to point out the principle that “the licensee retains the prime
responsibility for safety”. It is natural that the licensee owns the responsibility for safety
design and regular safety management. Emergency response is also an important role ofthe licensee. For example in Tokai Daini, when EDG-2C was stopped, operators
checked the situation of other pumps by paying attention to the tsunami first. Then they
prevented the extension of hazards, and decided the restoration procedure considering
the risk of each activity. These activities were performed in other facilities as well.
Workers on the site know the situation best, and the confirming the situation at the siteshould be the basis for the all types of emergency response. If emergency progressed
and reached to the higher level of DiD, the role of management became bigger.
Even in the beyond design basis situation, thorough understanding of thesituation on the site and responding appropriately to the situation is essential to preventthe situation from progressing, to mitigate the effect, and to settle the incident. These
problems are discussed in Section 7.2.
143

(Drafter: Yasuhiro Fujii, Kiyoharu Abe)

Reference
6-1
Tokyo Electric Power Company, “Fukushima Nuclear Accidents Investigation
Report”, June 20, 2012.
http://www.tepco.co.jp/en/press/corp-com/release/2012/1205638_1870.html
6-2
K. Nagasawa, “Situation at Fukushima Daini Nuclear Power Station”, Lectureslides for the 7th Seminar
6-3
Tohoku Electric Power Company, “Situation at Onagawa Nuclear Power

6-4

6-5

6-6

Station regarding to the Tohoku-Pacific Ocean Earthquake and Subsequent
Tsunami”, Press Release at May 30, 2011, in Japanese.
http://www.tohoku-epco.co.jp/news/atom/1183294_1065.html
Tohoku Electric Power Company, “Situation and Safety Measures at Onagawa
Nuclear Power Station after the Tohoku-Pacific Ocean Earthquake “, Lectureslides for the 6th Seminar, October 22, 2012, in Japanese.
The Japan Atomic Energy Company, “Analysis of Seismic Records andInvestigation of Tsunami at Tokai No. 2 Power Station regarding to the 2011
Tohoku-Pacific Ocean Earthquake”, May 2013, in Japanese.
The Japan Atomic Energy Company, “Situation at Tokai No. 2 Power Station
after the Tohoku-Pacific Ocean Earthquake”, Lecture slides for the 6th Seminar
October 22, 2012, in Japanese.

144

Table 6.4-1 Comparison of the maximum acceleration based on observed records and the maximum response acceleration
with respect to the reference ground motion
Unit: Gal
Observation Point: Basement of Reactor Building
Site

Fukushima
DaiichiOperation Status

S-N

E-W

U-D

S-N

E-W

U-D

Unit 1

Full Power Operation

460※１

447※１

258※１

487

489

412

Unit 2

Full Power Operation

348※１

550※１

302※１

441

438

420

Unit 3Full Power Operation

322※１

507※１

231※１

449

441

429

Cold shutdown(Regulatory Inspection)
Cold shutdown(Regulatory Inspection)
Cold shutdown(Regulatory Inspection)

281※１

319※１

200※１

447

445

422

311※１

548※１

256※１

452

452

427

298※１

444※１

244

445

448

415

Unit 1

Full Power Operation

254

230※１

305

434

434

512

Unit 2

Full Power Operation

243

196※１

232※１

428

429

504

Unit 3Full Power Operation

277※１

216※１

208※１

428

430

504

Unit 4

Full Power Operation

210※１

205※１

288※１

415

415

504

Unit 1

Full Power Operation

540

587

439

532

529

451

Unit 2Start-up Operation(Regulatory Inspection)

607

461

389

594

572

490

Unit 3Full Power Operation

573

458

321

512

497

476

Full Power Operation

214

225

189

393

400

456

Unit 4

Unit 6

Onagawa

Design Basis Seismic Motion Ss

Unit

Unit 5

Fukushima
Daini

Maximum Acceleration Observed by the Recorder

Tokai Daini※1: Record was stopped after 130 ~ 150 s
145Table 6.4-2 Tsunami in Fukushima, Onagawa, and Tokai
Site

Fukushima Daiichi※１

Elevation of main buildings

Unit 1~4: O.P.+10m

area

Unit 5, 6: O.P.+13m

Postulated tsunami height in
the application for reactor
establishment permit

Actions after JSCE method
was issued in 2002

Fukushima Daini
O.P. +12 m

O.P.+3.122 m
O.P.+3.122 m(for Unit 1 in 1966)(for Unit 1 in 1972)
O.P.+3.705 m(for Unit 3, 4 in 1978)

O.P. +5.7 m
Wave source: Off Fukushima
Seawater pumps were lifted to thehigher position, etc.

O.P. +5.2 m
Tightened the water proof of thebuildings, etc.

Onagawa

Tokai Daini※1

O.P. +14.8 m

H.P.+8.9m※2

O.P.+2～3m (for Unit 1 in 1970,
based on literature review)
O.P.+9.1m (for Unit 2 in 1987,
based on simulation)

H.P.+2.35m (in 1971)

O.P. +13.6 m
Wave source: Off Sanriku

H.P. +5.75 m

Countermeasure was unnecessaryCountermeasure was unnecessary
H.P. +6.61 m

O.P. +4.7 m

O.P. +4.7 m-

assessment in 2007

Countermeasure was unnecessaryCountermeasure was unnecessary-

Actions after Fukushima

About O.P. +5 m

O.P. +5 m--

Prec.’s assessment in 2007

Countermeasure was unnecessaryCountermeasure was unnecessary--

Reevaluation of submarine

O. P. +6.1 m

O.P. +5.0 m--

topography and tide

Seawater pumps were lifted to theconditionhigher position

Countermeasure was unnecessary--

Actions after Ibaraki Prec.’s

The highest observation at
the Tohoku-Pacific Ocean
Earthquake

Tsunami height: O.P. +13.1 m

Tsunami height: O.P. +7~8 m

Upstream: O.P. +15.5 m

Upstream: O.P. +14.5 m

Tsunami height: O.P. + 13.0 m

Wall around the seawater pumpareas were extended, etc.

Tsunami height: H.P. +5.5 m
Upstream: H.P. +6.2 m※1 O.P. ±0.00m for Onagawa is the standard level for Onagawa construction, and is 0.74 m lower than Tokyo Bay Mean Tide Level, i.e. O.P. ±0.00 m is T.P. -0.74 m
O.P. ±0.00m for Fukushima Daiichi and Daini is the standard level for Onahama Point, and is 0.727 m lower than Tokyo Bay Mean Tide Level, i.e. O.P. ±0.00 m is T.P. -0.727 m※2 H.P. ±0.00m is the standard level for Hitachi Port, and is 0.74 m lower than Tokyo Bay Mean Tide Level, i.e. O.P. ±0.00 m is T.P. -0.74 m

146

Table 6.4-3 Damage condition caused by earthquake and tsunami in Fukushima Daiichi, Fukushima Daini, Onagawa and Tokai Daini(○: available, ×: failure, [number of available/ total number]: partly available)
Site

DaiichiP/C（Low Vol. Board）(( ): under maintenance)
Emergency
Regular

×

×

×

×

×

×

2

*3

×

×

×

×

2/3

2/4

*3

○→dried off

×

×

×

×

×

3

×
275kV ×66 kV ×(Total 7 lines)

×

Partly Used

*3

×

×

×

×

1/2(1)

1/1(1)

*3

○

×

×

×

×

2/7

6

1/3

*3

○

×

○

×

○

×

1

*4

3/4

×

1/3

○

1/4

○

*4

×

○

×

○

○

2/4

○

2/3*4

○

○

○

3/4

○

1/3*4

○

○

○

2/4

○

○

○

○

1/2

○

○

○

○

○

○

○

○

○

○

○

○

○

4

3

2

×
×
×

500 kV: 1/2
66 kV ×(Total 4 lines)

1
275 kV: 1/4
66 kV ×(Total 5 lines)

3
Tokai No. 2

M/C（High Vol. Board）(( ):under maintenance)
Emergency
Regular×*3

4

Onagawa

Seawater
Circulation*7

1

2
Daini

Power supplycar

DC power

5

Fukushima

*2

EDG

Unit

Fukushima

External
Power*1

275 kV ×
154 kV ×(Total 3 lines)

1/3*5

○

○

○

Partly Used
(External
power, EDG
were secured)○for RHR: 1/2
for HPCS: ○for RHR: ×○(External
power, EDG
were secured)for HPCS: ×

for RHR: 1/2○
Backup2/3*6

for HPCS: ○(EDG weresecured)for EDG: 2/3
for RHR: ○Severe Core
DamageDamagedSecuredSecuredSecuredSecured*1: External power supply was restored in one ~ several days in Fukushima Daini and Onagawa. A 66 kV line in Fukushima Daini was under maintenance.
*2: No EDG was damaged by seismic motion.
*3: One EDG in Unit 1, 4, and two EDGs in Unit 5 was not failed by submerge. The failure was due to their relative components such as coolant system and M/C. EDG-Bs in Unit 2, 4, and 6 were air-cooled.
*4: Water flew through the suction port of EDG fans in Unit 1. Flooding in Unit 2~4 were very small. EDGs in Unit 2~4 failed due to the submerge of their relative components, such as coolant system and M/C*5: System A was available. System B failed due to the loss of reactor auxiliary coolant pump and loss of HPCS auxiliary coolant pump. They were submerged by the flooding from the intake of tide gage inseawater pump room.
*6: Seawater pump 2C for EDG-2C was stopped automatically due to the flooding in seawater pump room. EDG-2C was stopped manually.
*7: Seawater systems were failure. They are equipped in heat exchange buildings in front of the coast in Fukushima Daini, pit caved from the main buildings height in Onagwa, and seawater pump room whose wall
was lifted in front of the coast in Tokai Daini.

147

7.
7.1
7.1.1

Issues Identified by the Accident
Summary of issues According to DiD Concept
Relationship between Safety Measures and the DiD
The primary objective of ensuring safety is to protect the public in the vicinityof nuclear facilities from radiation harm. The fundamental concept used to realize
nuclear safety is known as the “defense in depth (DiD).” The point of the DiD is firstly
to provide multiple levels of safety measures, and secondly secures the safety by theeach individual level of the defence. When a safety measures at each level is examined,
the good effect of the previous level and the subsequent level should be neglected. To be
precise, measures at each level should be sufficiently reliable that it is not necessary to
relate them to measures at subsequent levels; and this is therefore considered to be a“negation of subsequent levels”. With such robust measures however, it is necessary to
prepare safety measures at all subsequent levels with the assumption that any precedingmeasures could fail; and this is also considered to be a “negation of preceding levels”.
If this understanding is expressed using PRA wording, DiD is then described asfollows. Multiple levels of safety measures are provided, the safety measures at eachlevel have sufficient reliability so that it is not necessary to relate them to measures on
preceding or succeeding levels. As a result, highly reliable safety is ensured throughout
all developments.
DiD has been used as the basic safety principle since the beginning of nuclear
power technology development, and technically achieving this principle is the goal ofsafety research and design processes. DiD for a nuclear power plant is generally
structured on five levels, as follows.
Level 1: Prevention of abnormal operation and failures
Level 2: Prevention of escalation of abnormal operation and failures in relationto accidents
Level 3: Mitigation of accident consequences
Level 4: On-site response to beyond design basis accidents
Level 5: Off-site response to beyond design basis accidents
The word “accident” used in the explanation above means a DBA. Accident conditions
that are more severe than DBAs are known as Beyond-DBAs, and include accidents
without core degradation and severe accidents in which core degradation occurs. Level
3 is categorized as pertaining to measures to be taken before a severe accident148

(including some D-DBAs) by some people, and Level 4 can be classified as measures tobe used in a severe accident. In any case, Level 4 relates to conditions that occur beyond
a DBA, and accident management (AM) is to be used as the major countermeasure.
Off-site responses involving AM, such as the delivery of various resources such as
equipment, materials, and preparation of portable safety equipment (which was added
after the Fukushima Daiichi accident), should be categorized as Level 4. Level 5 is an
emergency response involving the public.
Safety based on DiD is not only required in facility design, but should also be
included in all appropriate management and operations. A PRA is used to fully confirm
whether safety exists or not. In some cases, discussions have taken place in relation tochoosing between DiD and the application for the results of risk analysis. However,
PRA is intended as a method to measure how well DiD is being adequately
accomplished. If any small risks are identified then reasonable efforts should be made to
reduce risk and enhance safety, such as making improvements in facilities andoperational procedures.
If the failures of SSCs occur randomly by miscellaneous initiators, afive-layered DiD could therefore be successfully applied to nuclear power plants byinstalling AM equipment, in addition to making safety equipment diverse and redundant.
In the Fukushima Daiichi accident however, many SSCs failed in relation to a common
cause (seismic motion and the tsunami), which exceeded the design postulation.
Consequently, the reactor core melted down and a large amount of radioactive materials
were released into the environment. This raised the question of whether DiD against
external initiators, particularly against natural events, had been sufficiently implemented
in nuclear power plants.
7.1.2

Particularly prevalent problems in the Fukushima Daiichi accident
Problems exposed in the Fukushima Daiichi accident are related to all five levelsof DiD, and these are summarized as follows, according to layers.
The objective of the first level is to prevent the occurrence of anomalies
including initiating events and failures. Providing sufficient margins in safety design
utilizing proven technologies, providing appropriate protection when necessary against
specific initiators such as seismic motion and airplane crash and conducting
maintenance activities in accordance with the quality management system are includedin this layer. In the Fukushima Daiichi accident however, a large amount of safety
equipment simultaneously lost its function as a result of the single external initiator (the149tsunami wave).
The objective of the second level is to prevent the escalation of incident by
responding as soon as a problem is diagnosed. For example, control rods areautomatically inserted to shut down the reactor when operation parameters exceed theallowable range. So far at this level no problems have been isolated from theinformation obtained. Before the tsunami struck the site after the earthquake, thereactors shut down automatically due to the signal from the seismometer. Soon after theexternal power was lost, the EDG were activated. The reactor was therefore controlled
as expected in accordance with its design.
The objective of the third level is to mitigate the consequences of the accident,
in the case of its occurrence. For instance, the ECCS is provided to cope with a LCOA
where piping in the reactor’s coolant system is ruptured or leaked and the reactor
coolant is released. In addition, a robust containment vessel is used to prevent therelease of radioactive materials into the environment, and containment of the coolingsystem is intended to decrease internal pressure to retain the integrity of the containment
vessel. In the Fukushima Daiichi accident, however, the tsunami wave exceeded thepostulated height. AC power, the seawater cooling system, and DC power were lost,even though the tsunami wave affected the units slightly differently. Many safetysystems failed and the reactor core was therefore melted down. The regulatory attitude
against SBO was also highlighted after the accident because the "Regulatory Guide for
Reviewing Safety Design of Light Water Nuclear Power Reactor Facilities” 7-1 (Safety
Design Guide), issued by former NSC, stated that evaluation of the consequences oflong-term SBO is unnecessary.
The fourth level pertains to AM measures, and these have been prepared in twophases. The first phase AM measures are prepared against accidents that occur beyond
DBA to prevent them escalating up to a severe accident. The second phase AM
measures are prepared to mitigate the effect of severe accidents in the case that an
accident escalates up to a severe accident. Some of the AM measures worked effectively
during the Fukushima Daiichi accident, but the other AM measures did not deliver
expected results because the earthquake and the tsunami damaged facilities both on-siteand off-site, and in addition the high radioactivity and other effects of the severe
accident prevented the initiation of these AM activities.
The fifth level relates to the emergency response off-site. It must be said thatthe effectiveness of emergency preparations was very low in Japan. A problem with thecollaboration between organizations has been highlighted. Moreover, the accident
showed that lessons from the JCO nuclear criticality accident at Tokai-mura in 1999

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(such as arrangements for people in need of care) had not been learned.
Although many problems were identified, the following three problems are
considered to be of particular importance:
1) Defenses against the external initiators, in particular, defenses against natural events,
were insufficient. The methodology to assess the tsunami height was not matuured.
Design basis hazard (DBH) must be determined so as to take sufficient enoughmargin against each external initiator, even considering the very rare events In
reality, however, adequate postulation was not made against tsunami.
2) The AM was not sufficiently reliable when encountering a real severe accident.
When the event occurred (that was beyond the design basis), the prevention ofevolution to severe accident and the mitigation of consequences should have been
made according to the AM measures. However, the effects to the work environment
from an earthquake and tsunami, as well as those from severe accident were notadequately considered in the preparation of the AM. As a result, implementation of
the AM under actual accident conditions was difficult.
3) “Flexible measures” against “beyond postulation events” were not considered. No
matter how well an event is discussed and considered beforehand, there always
exists a possibility that such safety measures are not functioning as expected.
Therefore, in order to avoid the worst situation, portable safety equipment shouldhave been prepared (for example). Some of these measures had been alreadyinstalled in the U. S., but there were none in Japan.
The following sections in Chapter 7 describe in detail the problems highlighted byreflections on the Fukushima Daiichi accident. These three main problems and theirsolutions are, however, summarized below:
1) Defenses against external initiators; in particular against natural phenomena
The first level of DiD is to prevent failure of the SSCs. The requirements on
SSC design are specified in the “Regulatory Guide for Reviewing Safety Design ofLight Water Nuclear Power Reactor Facilities” by former NSC (“Safety Design Guide”)
7-1. Its Chapter IV“General Requirements for Nuclear Reactor Facilities” has 10 specific
Guidelines (Guidelines 1 to 10) all of which require that safety related SSCs must havehigh reliabilities. Especially, the Guidelines 2 to 5 request sufficient consideration indesign against natural phenomena, external man-induced events, internal missiles, and151

fire, respectively.
An assessment of the height of a potential tsunami wave was discussed by
JSCE, and standards for this were issued 7-2. The design basis for the tsunami wave wasgiven by the Seismic Design Code and the Seismic Design Guide, both of which were
issued by The Japan Electric Association 7-3, 7-4. The assessment method used by JSCE
was referred in these standards when deciding the postulated tsunami height. Originally,
JSCE has established the standard for general civil engineering structures, and in thedesign processes they assumed that a tsunami would occur at a frequency of aroundonce per every 100 years. In contrast, nuclear designers considered an initiator with afrequency of approximately once per every 1000–10000 year as the design basis. Thereis therefore a difference in the practice used by the various engineering fields. The
collaboration and communication between academic societies was considered to be
insufficient.
Defense should be designed appropriately with respect to each potential
initiator. Against seismic motion, the seismic design of each SSC should be conductedwith the incorporation of adequate safety margins. Against tsunami waves, a seawall
incorporating factors such as adequate height and water-proof gates should be built tocontrol risk. PRA and stress tests can provide useful information in the design ofdefenses against initiators.
2) Reliability of AM
The fourth level of DiD consists of AM for severe accidents. In the Fukushima
Daiichi accident, much equipment failed in relation to a common cause; the earthquake
and tsunami. Some of the AM equipment was also damaged, which made AM activity
difficult. Damage occurred throughout the entire site. Both access to the facility and thedelivery of necessary resources were therefore difficult. In addition certain
communication tools failed, and as a result sharing of information and collaboration
were obstructed. It is considered therefore, that the possibility of natural events should
be assumed more widely, and efforts to reduce the impact of unexpected effect causedby the natural event should be conducted. When the AM measures relating to a natural
event are prepared, inherent obstacles that will potentially arise in relation to such an
event should be carefully considered.
3) Flexible measures against events that reach beyond postulation
The Fukushima Daiichi accident raised the question of what the appropriate
design of a DiD should be in relation to a huge natural hazard with a very low frequency.

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As already mentioned above, nuclear safety is achieved technically using the DiD
concept, and an adequate system against random failures has been already quite wellestablished. It is evident that the DiD design was insufficient against an external hazard
including natural phenomena, and this was related to the difficulty experienced inconsidering reasonable and effective measures for protecting the equipment on the site.
For example, there is a possibility that the facilities in the reactor building could
malfunction in relation to a common cause, such as if the earthquake was more extreme
than the one considered within the design postulation. In the case of the tsunami,
flooding far beyond design postulation led to the loss of safety functions, which would
have occurred even if the water-proofing of the reactor building had been enhanced.
Difficulty of safety measures against natural hazards is made by the existence of “thecliff edge” whose frequency is very low. Once the hazard of a natural event exceeded acertain level, so called a “cliff edge”, the consequence of the event can be huge.

(Drafter: Kiyoharu Abe, and Kumiaki Moriya)

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7.2

Issues on Safety Design
Accident management (AM) measures have been installed in Fukushima
Daiichi as explained in Chapter 3. The off-site power supplies to Fukushima Daiichi
were lost by the Tohoku-Pacific Ocean Earthquake. After 40 minutes from the main
shock, the power station was inundated by the tsunami, which led to blackout and lossof the ultimate heat sink. Consequently, it delayed both depressurization of reactors andcoolant injection into reactors and damaged the reactor core. Since loss of DC power
and failure of power distribution boards prevented all prepared AM activities fromfunctioning, water injection into both the reactor core and the PCV was delayed. Then,the core was damaged by loss of cooling, and the PCV was damaged by over-heating
due to the delay. As a result, these induced the release of radioactive materials andcaused soil contamination.
This section illustrates the lessons learned from the viewpoint of safety designof a nuclear power plant, including AM, related to the above accident responses.
7.2.1

Event Analysis of the Fukushima Daiichi Accident, and the Lessons

Related to Safety Design
It is important to identify lessons related to safety design against the external
initiator from the event analyses of the Fukushima Daiichi accident. We performed
event analyses as described in Figure 7.2-1~3 at the Seminar and obtained sevenlessons in plant design. The measures, which are described as an application of eachlesson, are presently considered to be effective. Specific consideration and more
improvement will be required when these lessons are technically applied to existing
plants.
(1) Lesson 1: Arrangement, backup, and recovery of important equipment
The beyond design basis tsunami, which flowed over the bank protection, left
the Fukushima Daiichi submerged beneath almost 5 m of water. This tsunami flooded
the first floor of the reactor building through air inlets and outlets and doors broken by
the impact force of water. Emergency AC and DC power supply system andswitchboard lost their function as flooded water entered the underground floor throughstairs.
Rearrangement of important equipment on a higher place and/or waterproofing
of important equipment and facilities are effective protection countermeasures.
On the other hand, backup provisions and recovery of important equipment are
effective mitigation countermeasures, because air inlets and outlets should be installed,154

since beyond design basis impact force of water can occur. Backup provisions of power
supply trucks, switchboards, and fire engines should be present outside the reactor
building as a hill or a waterproof backup building. It is also important to provide an
emergency connection for mobile equipment on reactor buildings and to have heavymachinery available for establishing access routes.
(2) Lesson 2: Configuration and deployment of isolation valves
The IC or the RCIC works as the core cooling function during isolated
conditions of the RPV. During loss of its function, low-pressure core injection should be
used after HPCI operation or RPV depressurization. In Fukushima Daiichi Unit 1,
isolation valves of the IC seemed to be closed due to loss of instrumentation DC power,
because DC power for the instrumentation was lost prior to AC power, which is thepower source of closing valves, during the large-scale tsunami.
Logic can be developed that prioritizes cooling considering severe accidents.
However, the isolation function should be prioritized higher, because boundaries should
be maintained for design basis faults and there might be a necessity of isolation during
severe accidents. Therefore, measures for opening closed isolation valves, which can be
operated outside the containment or remotely, should be added. It may be achieved byequipping isolation valves only outside of PCV, or by preparing backup DC power for
operation of the valves.
(3) Lesson 3: Provision of backup DC power supply
Loss of DC power caused the following situations during the Fukushima
accident: Difficulty in understanding the plant status due to loss of all instrumentation
functions, which caused delay of initial response and failure of appropriate recovery
actions. RPV depressurization with SRV was delayed because the valves of theaccumulator had to be opened by DC power. DC power was also needed for control
signal for the coolant injection systems such as RCIC.
Accordingly, it is important to prepare portable DC power and battery charger
to supply DC power for these important functions even though the regular system lose
its function. The capacity of the backup DC power can be smaller than the regular one.
It is rather important to consider the portability during the emergency.
(4) Lesson 4: Instrument reliability and credibility
The reliability, credibility, and/or continuity of the instrumentation of thereactor water level, pressure, and temperature, which are important for AM, were lost155

with progression of the accident. Since the environment and application ranges
exceeded the design condition largely, the credibility of the instrumentation was lost.
Since the continuous measurement was impossible during the accident, the whole trendof obtained data could not be examined, and it took time to realize that the credibility of
the instrumentation should be doubted.
Reliability of instrumentation used during a severe accident is less important
than those used for design basis. The instrumentations for severe accident should have ameasure for confirming credibility, because a drift or malfunction may occur under
severe conditions. Continuity of the instrumentation should be also improved bystrengthening power supply systems.
In addition, accident management procedures and practices for loss of all
instrumentation should be prepared.
(5) Lesson 5: Provision of a flexible coolant injection
Not only protection measures such as waterproofing and rearrangement forimportant equipment but also provisions of flexible coolant injection into RPV andcontainment are important. Considering an ultimate situation in which there is loss of allpermanently installed coolant injection systems, the coolant injection should bediversified by mobile equipment. The diverse water source should also be prepared,
considering the transportation of water from outside of the facility.
Flexible and simple management procedures with the mobile equipment areimportant, and the training utilizing these procedures should be conducted continuously.
(6) Lesson 6: Accessibility, operability, and effectivity of AM equipment
There were problems with the accessibility, operability, and assurance ofeffectiveness of AM equipment during the Fukushima accident. It was difficult for theoperators to access the S/C venting valve because it is located near S/C, which has noshield. Existence of the rapture disc, which was installed to prevent the loss of theboundary due to the wrong operation of the valves, obstructed the ventilation performed
quickly. The connection port of the alternative coolant injection from outside the reactor
was also difficult to be accessed. Even after the connection, effective injection was
delayed due to the existence of bypass of the water.
A remote manipulation handle can be added on the manually opened isolation
valves to reduce the radiation exposure. Connection ports for alternative coolant
injection can be equipped distributedly around the reactor building, so that the operators
can access to a connection port easily.

156

To improve the operability of the AM for cooling function, isolation valves canbe additionally installed between the line shared by AM equipment and the normal load,so that the bypass of the water can be avoid. Installation of coolant injection line for
SFP also improves the operability.
To improve the effectiveness of PCV venting, the role of the rapture disk should
be reconsidered. It is necessary to add new line to bypass the rapture disk, or to remove
the rapture disk.
(7) Lesson 7: Provision of alternative means for protecting containment vessel
A major reason of land contamination by the Fukushima accident is assumed to
be a leakage from the containment gasket, because of the over-heating of thecontainment following a delay of core debris cooling. On the other hand, fission productrelease through W/W venting with scrubbing is assumed to be much less than those by
the leakage from the gasket. Therefore, a countermeasure for over-heating of thecontainment, that is for cooling the core debris, is important.
To avoid the leakage due to over-heating, two countermeasures are considered:
1) To avoid the containment over-heating sequence, the coolant injection ability into
RPV and containment can be enhanced. In this case, W/W venting should be adopted tocounter the over-pressure sequence; 2) To improve the safety margin against
over-heating, non-metallic parts of a containment head flange can be cooled with adirect coolant injection or by storing water in the well of the reactor.
7.2.2
Summary of Issues and Measures, Concerning Safety Design on theConcept DiD(1) Summary of issues
Three issues were pointed out when we considered safety design in theperspective of DiD described in Section 7.1.
The first issue is that protection was insufficient for external initiators, inparticular natural events. Unlike random failure events, external initiators can cause thecommon failure of redundant SSCs. Both the break of the first defense level of DiD andthe break of subsequent defense levels of DiD may be caused simultaneously by single
external impact from the viewpoint of physical hardware. In this manner, there is apossibility that a lot of equipment lose function. Such characteristics had beenrecognized in the safety research before the Fukushima Daiichi accident; however,appropriate and effective measures after the first protection line failed had not been157

developed. The only effort invested has been to improve the margins, such as seismic
design and the height of tide walls. We should have established effective
countermeasures for subsequent levels of DiD.
The second issue is the difficulty to set up protection levels against natural
events. Hazard evaluations used in setting up protection levels for natural events aremainly based on historical experience data. Although modelling for natural events andsimulation technologies are applied in modern ways, the design of protection levels
have been based on historical data. Then, historical data has limitation in period andlarge uncertainty in these records. Therefore, we should have a larger recognition of thisuncertainty and the event that exceeds assumption.
With regard to the points mentioned above, large margins have been set for
design basis hazards in accordance with the DiD principle. This approach worked
effectively against the earthquake at the Fukushima Daiichi accident. The nuclear
reactor equipment are classified according to their importance in terms of safety.
Important equipment are designed with larger margins. With the study updated on theearthquake, a back check was performed and proper and effective effort was made toreflect the new understanding. Even though there might be room for improvement on
quick response to the new information, protective measures against seismic motion havebeen effective.
In terms of the tsunami, the protective measures were designed based on theconcept that ‘tsunami should be stopped by the height of the site and tide wall’.
Postulated tsunami height had also been re-assessed following to the updated study, but
was extremely exceeded in the Fukushima Daiichi accident. With respect to issues on
assessment of the tsunami height, as discussed in Section 7.1, margins for the tsunami
were not taken into account as for earthquake. Furthermore, countermeasures against
unexpected tsunami should also have been seriously considered though the latest
findings were included in tsunami assessment. We should have recognized the difficultyof assuming hazards of multiple disasters such as tsunami in association withearthquake. Damage to the plant and the site is difficult to define when the hazard
exceeds design assumption. Effective measures should be developed while considering
these difficulties.
The third issue is the difficulty of carrying out AM as indicated in Section 7.1.
It is necessary to take measures in order to rapidly carry out appropriate AM even if theentire plant was affected by disasters.

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(2) Summary of response measures
Considering the issues described above, response measures are discussed asfollows:
To solve the first and second issues, we should not simplify our viewpoint
focusing only on the equipment design in the plant and at the site. Larger systems
including both on-site and off-site should be taken into account.
Not limited to measures on-site, proper measures in a large system including
off-site systems should be considered. If measures against external initiator, in particularnatural events, were only installed in the plant, equipment in the reactor building canlose their function according to the common cause on the occasion of a hazard far
beyond postulation. Hence, preparing AM at off-site taking into account the charactersof external initiators that exceed design load is reasonable in order to avoid the risk offailure of equipment due to common causes.
In response to the Fukushima Daiichi accident, fire engines with a hoseconnection as AM lines were effective to inject water during the situation where almost
pre-existing equipment were losing their function. This fact enhanced the effectivenessof portable equipment. Besides, fire engines were effective even though they are notdesigned in accordance with nuclear grade, which requires high reliability. It shows thatpriority should be given to flexibility rather than high reliability and high performance.
These ideas went up for discussion before the Fukushima Daiichi accident but were
postponed due to the limitation of adaptability to the nuclear grade specification. This
proves the necessity of reasonable evaluation of equipment ability. Not only redundancy
and robustness but also diversity and flexibility should be valued in the phase ofevaluation including regulatory evaluation.
Regarding the third issue, various unexpected events should be assumed, andmeasures for effective implementation of AM should be considered. Preparation of
hardware, such as ensuring access routes towards AM equipment and disposing
necessary heavy machines, is of course important. Preparation of software, such as
preparing procedures for accurate judgment and execution for unexpected events,
education, training, preparing operational structures, and so on, are also very important.
Finally, the Fukushima Daiichi accident was an appearance of the “cliff edge”,
which has been pointed out in safety research for a long time. Even though its frequency

159

was negligible in the engineering view, a reasonable safety net should be spread in
accordance with its consequence. The accident clearly shows that the reasonable safety
net should not be limited by on-site measures. The safety net should be installed outsideof the plant as well, while paying attention to the diversity of location that involves bothon- and off-site activity. Response measures should be equipped with reasonable
reliability, which should be decided according to the frequency of the target. The
accident showed that excessive redundancy is unnecessary. It is more important to findthe best collaboration of software and hardware than making equipment redundant.
Finally, improvements should be made soon after they are considered necessary. A
system for adopting countermeasures immediately is required.
(Drafter: Kumiaki Moriya)

160

Fig. 7.2-1 Lessons Learned and Countermeasures (1) Fukushima Daiichi Unit 1
161

Fig. 7.2-2 Lessons Learned and Countermeasures (2) Fukushima Daiichi Unit 2
162

Fig. 7.2-3 Lessons Learned and Countermeasures (3) Fukushima Daiichi Unit 3
163

7.3
7.3.1

Issues in Safety Regulations
Reconsideration of Principles in the Light of the Accident
The Fukushima Daiichi accident revealed that many problems also exist in
relation to nuclear regulation. The Nuclear Regulation Authority (NRA) and its
secretariat agency were subsequently launched on September 19, 2012 and new
regulations were initiated by the NRA. The NRA is now expected to improve
regulations that reflect the lessons learnt from the accident. In this section, we identify
what were improper or insufficient in the past regulation, and show future expectations
for the improvement of regulation.
In this Section, Clause 7.3.2 recounts the problem of the delay in “continuous
improvement” within the regulation. Insufficient consideration of natural phenomena
and the delay of regulating severe accidents are two of the most important problems,
and these are taken up in Clauses 7.3.3 and 7.3.4, respectively.
We also discuss expected improvements in regulation. However, we firstly wantto emphasize that the general safety principles that existed before the accident are still
important now. The most typical example of this is the importance of DiD, which has
been re-recognized as being important internationally, and this has been concluded in
Section 7.1 within this report. Two other important concepts are chosen in this section.
Importance of scientific and reasonable judgments on regulation is taken up in Clause
7.3.5. We discussed how to achieve both independency of regulation and collaborationof all industries, government, and academia in Clause 7.3.6.
7.3.2

Importance of Continuous Improvement within the Regulation
As described in Section 6.1, the Fukushima Daiichi accident was caused by thetsunami that occurred after the seismic motion. In relation to tsunami countermeasures,
licensees had somehow promoted continuous improvement which to a certain extent
reflects new knowledge in that area. The tsunami height hypothesized within the design
had been gradually raised, reflecting newly obtained knowledge, and the tsunami height
estimation methodologies and anti-tsunami design had been enhanced according to theraised tsunami height. It is still the responsibility of operators to carry out actions, as“the prime responsibility for safety must rest with the operators.”
The accident revealed that the operators’ activities were not sufficient at
Fukushima Daiichi, and if we evaluate other nuclear power stations that were affectedby the tsunami, it is clear that severe accident was narrowly prevented as a result of the
improved activities of the operators therein.
Continuous improvement is also essential in relation to the role of regulatory164

bodies. A principle exists that regulation should be based on the most up-to-dateknowledge, and therefore a framework for periodic safety reviews (PSR) has been
prepared. Existing plants are required to respond to any new knowledge, particularly in
relation to the revision of standards and guides.
In relation to tsunami countermeasures, however, the previous NSC design
requirement was very short and was merely attached at the end of the NSC Seismic
Design Guide. Regulatory bodies endorsed technical code (JEAC 4206) 7-3 and technicalguideline (JEAG 4206) 7-4 for seismic design, both issued by Japan Electric Association.
In the code and the guide, latest assessment guidelines issued by JSCE 7-2 was only
referred. As the result, the basic concept that was used to set postulated tsunami wave
did not correspond to the objective of nuclear safety.
In relation to severe accident countermeasures, all the nuclear power plants inJapan prepared accident management (AM) in the 1990s, as a voluntary activity of eachutility. The regulatory body at that time considered that the next step would be toestablish regulatory requirements on severe accident countermeasures; however thisregulatory action was not carried out.
The feedback of operational experiences was also insufficient. There was ageneral understanding of the importance of reflection on past operational experiences
within Japan and overseas. The flooding event which occurred at Madras Atomic Power
Station in India caused by Great Tsunami in the Indian Ocean was known not only by
the nuclear industry but by the regulatory organizations in Japan. However, it was
regarded as being “just an event that occurred in India and is not relevant in Japan.”
Neither was the regulations influenced by the September 11 terrorist attacks. Japan’s
regulatory bodies had to learn from these experiences seriously.
From now on, improvements based on reflection on the Fukushima Daiichi
accident should commence immediately. In order for this to occur, it is necessary tounderstand what really happened during the accident (See Section 7.6) to identify whatwent wrong, and then to promote improvements corresponding to the identified
weaknesses. NRA has already started reviewing design bases concerning earthquakes,
tsunamis, and severe accidents. Back-fitting will then be required at all existing nuclear
facilities in relation to these new safety requirements.
Although we regret the occurrence of the Fukushima Daiichi accident, it does
not automatically mean that we have to adopt all the proposed measures that come tomind. Reasonable improvements for reducing risks need to be implemented, as
described later in Clause 7.3.5.

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The approach used in implementing improvements should also conform tointernational concepts. The IAEA’s backfit standard described in Specific Safety
Requirement No. SSR-2/1 is shown as follows. “The design of new nuclear power
plants now explicitly includes the consideration of severe accident scenarios andstrategies for their management.” “It might not be practicable to apply all therequirements of this Safety Requirements publication to nuclear power plants that are
already in operation or under construction; in addition, it might not be feasible tomodify designs that have already been approved by regulatory bodies. For the safety
analysis of such designs, it is expected that a comparison will be made with the current
standards, for example as part of the periodic safety review for the plant, to determine
whether the safe operation of the plant could be further enhanced by means ofreasonably practicable safety improvements.” 7-5
7.3.3
Issues Involved in the Design Basis related to Initiators such as Natural
Phenomena
The Fukushima Daiichi accident showed that the design basis for tsunamis, oneof several hazardous natural phenomena, had so far been underestimated. The past
design and regulations had not sufficiently considered effective and concrete
countermeasures against specific initiator except seismic motion. A typical example isthe insufficient consideration involved in defining DBH for some natural phenomena.
At the beginning of the Safety Design Guide issued by the former NSC, aseries of requirements for countermeasures against initiators are listed. It states that SSC
that have a safety function should be designed in consideration of their ability towithstand seismic motions. The Guide also requires that the “safety function shall not belost under a postulated natural phenomena other than seismic motions.” “SSCs with thehighest importance of safety shall be designed in consideration of the most severe
natural phenomena or superposition of the natural event and the accidental load.” In
addition, the supplemental description within the Guide says that “the most severe
natural phenomena should not be less than the largest past record of a natural
phenomenon that is considered to be statistically adequate.” 7-6
This description could affect the quantification of some DBHs. When therecord of an initiator was limited and/or its statistical reliability was not sure, DBH
regarding to the initiator was decided based on the records in short period of time.
In the case of a tsunami, nuclear utilities re-evaluated the postulated tsunami
using the assessment method issued by the JSCE in February of 2002 7-2, and improved166

measures for coping with a newly postulated tsunami. The JSCE’s method was based on
the reliable historical tsunami records, from the period 1611 to 1978, and this method
itself was very consistent with the supplemental description in the Safety Design Guide.
The postulated tsunami quantified using this method appears to correspond to thehighest historical tsunami from the period 1611 to 1978, which means that the highest
tsunami occurring over a period of about four hundred years was selected for use in thesafety design. Even if the safety margin was considered, the frequency of occurrence thepostulated tsunami wave was about once in every thousand years (10 -3/year) and nocountermeasures were made against a tsunami that was higher than that postulated. It
was obviously impossible to satisfy the safety performance objective in Japan, that is 10-4/year of core damage frequency (CDF) and 10 -5/year of containment failure frequency(CFF).
It is considered to be quite natural for the JSCE to develop the evaluation
method of the tsunami height based on historical tsunami records. However, it seems
that neither the JSCE nor any other body has discussed the meaning of the use of thehistorical tsunami records in safety design from the view point of nuclear safety. A lackof communication between technical fields is to blame for this. Responsibility of thecommunication should have mainly rested on the nuclear engineers who manage therisk of the industry. It is therefore considered necessary to enhance smoothcommunication between various bodies involved.
The NRA is preparing clear regulations in relation to each individual initiatorthat encompasses a safety significance. The following two-step approaches are required:
1) the DBH should be set; 2) an appropriate protective design against DBH, such as asea wall against tsunamis, should be provided. Throughout the phase of deciding on the
DBH, it should be remembered that records such as those pertaining to historicalearthquakes and tsunamis are only relevant to a certain period of time. If we consider
the maximum hazard occurring over a period longer than that of the historical record
period to define the DBH, this should be evaluated by expert judgment.
However, even if DBHs are re-defined in such a manner, there is a possibility
that a beyond design basis hazard (B-DBH) can occur. In fact, neither the occurrence ofan earthquake of such a large magnitude, nor the overlapping of tsunamis fromsynchronized seismic sources had been widely recognized before the Tohoku-Pacific
Ocean Earthquake. In addition to the two steps related to the regulatory activities
mentioned above: 3) “Being prepared for the occurrence of an event that exceeds design
postulations,” should also be included by the regulatory body.

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There also was a problem in the NSC’s “Safety Design Guide” in relation toprotective measures against a DBA, which corresponds to the third level of DiD. In
relation to the assumption of a DBA, the Guide mentioned clearly that “there is no needto consider long term loss of AC power.”7-7 In the Fukushima Daiichi accident however,the station blackout continued for an extremely long period of time. Such inadequate
guidance was also due to simple reliance on past blackout records. The guide for SBO
was inadequate, and it was established based on the insufficient experiences.
In the Integrated Regulatory Review Service (IRRS) of the IAEA to Japan in
7-82007 , the former NISA received the recommendation that “As the regulatory body inJapan, NISA should take major responsibility in the development and endorsement ofsafety regulations and guides.” 7-9 This recommendation is very poignant in light of theevents that occurred in the Fukushima Daiichi accident.
Basic Design had been reviewed by the Safety Review Guides of the former
NSC. NISA did not settle on the regulatory basis by itself (with a few exceptions suchas “NISA’s Guideline to Assess the Frequency of Airplanes Crashing into Nuclear
Power Plants”). The resources and systems employed to continuously review the systemof regulatory guides were insufficient.
Standards for regulatory reviews need to be continually reviewed and updated,
and the creation of a responsible organization to conduct reviews and updating isnecessary. The NRA has already established a responsible division so called“Regulatory Standard and Research Division,” to review its regulations. This division is
expected to attempt the continuous improvement of the regulatory basis, by continuallyobserving the entire regulations from various perspectives.
7.3.4
Establishment of Regulatory Requirements against Accident Management
MeasuresEstablishment of regulatory requirements against severe accident measures is an
example of a delay in correspondence. In the 1990’s, inclusive AM was installed in all
the light water reactors in Japan, and this was not late compared to the world trend atthat time. It was therefore considered that the establishment of the regulatory
requirements of AM would be the next process. However, a review of whether the AM
equipment functioned effectively during a real accident did not become the object ofregulation. It is considered indispensable to make severe accident measures theimmediate object of regulations.

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It is of particular note that there were many events which were described as“beyond postulation” during the Fukushima Daiichi accident. It is necessary to consider
all risk that can be imagined, and appropriate approach against each imagination should
be taken. Therefore, a change in the design basis is definitely required when theimagination is insufficient. It is also necessary to imagine what could happen if thedesign basis was exceeded. However, problems occurred in the events that were
extremely difficult to imagine happening, such as the hydrogen explosion in the reactor
buildings and the explosion at Unit 4 that occurred in relation to hydrogen flowing infrom Unit 3. In expecting an occurrence that is impossible to imagine, flexible
countermeasures should be considered.
As described in Clause 7.3.2, measures were taken with respect to formulating
measures against an event that was beyond reasonable postulation after thesynchronized terrorist attacks using airplane in the US, but this was not reflected in any
regulations in Japan. In consideration of man-induced external initiator such as aterrorist attack, plant owners should prepare means of flexible correspondence withpersonnel and resources outside the site, and the regulatory body should review thevalidity of these measures.
7.3.5

Regulations using a Scientific Rationality are still Necessary
After the Fukushima Daiichi accident, a lot of people in Japan emotionally focus
on tightening the regulation. But regulation should be generally based on a scientificrationality. The use of risk information, that is compiled as a result of a probabilistic risk
assessment (PRA) and is therefore known as a risk-informed regulation, is
indispensable for this purpose. The attitude of apportioning resources based on the sizeof the risk, a so-called graded approach, needs to be adopted by both plant owners andthe regulatory body. Regulatory basis based on performance objectives is also aninvariant principle. In doing so, regulatory body only decides the items required.
Acceptable approaches for meeting the requirements should be described in the codesand standards issued by academic societies, and these are reviewed by the regulatory
body.
Before the Fukushima Daiichi accident however, the use of risk information
relating to regulations was performed only in relation to “the Revision of Seismic
Design Guide,” and was only used to “enforce the requirement for seismic design
because the risk of earthquake was high”; which was not sufficient enough reflection ofrisk information.

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There was a large difference between the effects of the seismic motion and thatof the tsunami during the Fukushima Daiichi accident. Seismic motion exceeded thedesign postulation in part, but was almost in the range of the design basis. All
Fukushima Daiichi Unit 5 and 6, Fukushima Daini, Onagawa, and Tokai No. 2 did notexperience failure of safety important SSCs due to seismic loads. In the case of
Fukushima Daiichi Unit 1~4, the possibility of the failure caused by seismic motion isconsidered to be very low though the certain evidence has not obtained yet. However,
the occurrence of seismic motion beyond the design basis was itself a problem, because
Guide 2 of the Design Guides mentions that the “design basis should not be less than
the reliable records of a natural event which is statistically adequate.” 7-6 However, theseismic motion did not cause severe effects as a result of the reconsiderations of designbasis and implementations of the plant. In contrast, the damage caused by the tsunami
was huge because insufficient considerations related to tsunamis had been implemented.
The current PRA is only performed in relation to two initiators, random eventsand seismic motion. The basis for a hazard evaluation was prepared in relation to asituation where an airplane crashed into the plant, to judge whether the collision
frequency would be high enough to require specific protection. There is no other PRA
prepared in specific guides and/or standards except for the one mentioned above.
Standards for tsunamis, fires, and terrorist attacks are lacking and/or are insufficient. An
individual plant examination of external events (IPEEE) was therefore needed toascertain the weaknesses of each plant, and this had not been carried out.
It is evident that, in both the regulations and in safety research, the object ofthese activities was focused on what was well known/understood by related personnel.
It is considered more important to find hidden problems not previously realized, and to
focus on such problems. For example, safety measure against intruders was notinvestigated thoroughly, even though many people knew that it is important.
The problems in “application of safety goals to regulations” also led to a delay
in applying risk informed regulation. Safety goals are initially needed to make rational
regulation. It can be said that nothing else relating to the improvement of regulations
was performed, in particular regulatory rules, except for the enforcement of the Seismic
Design Guide. The proposed use of safety goals was far different from the concept andreality of the PRA, such as “to perform a good or bad evaluation by comparing safety
goals and the results of PRA.” This is another example of the delay in improving

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regulations.
A review of the indexes of the safety goals may also be necessary. Current
indexes only relate to human health. Although nobody was killed by a radiation-relateddisease in the Fukushima Daiichi accident, the land was unfortunately highly polluted.
It is therefore necessary to add "the prevention of the soil contamination” to the safety
goals.
This is originally written in the interim report of Safety Goals TechnicalCommittee as follows: “There are various effects from radiation events such asindividual acute and latent death, the number of acute and/or latent death of the
population, and economic losses. Of these risks, the Safety Goals considers only the riskof individual death. Therefore, the Safety Goals do not cover all the risks but took up
only the most important risk, which can be quantified. In this context, the proposed
7-10
Safety Goals in this report are the first step of the beginning”.
However, the second
step of further setting safety goals was not conducted.
In the Eighth Seminar, Dr. Fuketa, the Commissioner of NRA, discussed Safety
Goals as follows: 7-11
“In relation to “residual risk,” in other words “safety goals,” I think that concrete
concepts regarding risks that remain after all possible countermeasures have been
employed, are important. The last stronghold of safety measures may be naturally
designed based on the deep consideration about residual risk.” “For safety goals, I
looked at global regulation practices and realized that approaches differ between
countries. Although the NRA was established three month ago, discussions related tothese approaches have not yet been concluded. Safety goals must firstly be discussed
and then a consensus must be reached.” 7-12
It is expected that these discussions will continue at the NRA and that risk
informed regulations will be promoted in line with the safety goals.
7.3.6
Issues Relating to the Independency of Regulations versus Collaboration
between Industries, Academia, and the Government
The IRRS to Japan, which was described in Clause 7.3.3, made a specificrecommendation as follows: “NISA is effectively independent from (Agency for Natural
Resources and Energy) […]. This situation could be reflected in the legislation moreclearly in future.” 7-13 This recommendation was automatically achieved when the NRA
was established in September, 2012.
After the Fukushima Daiichi Accident, some believe that regulatory bodies
should have no link with industries. It is considered, however, that the independency of171regulations should be consistent with the collaboration between industries, academia
and government.
During the Eighth Seminar, a participant made a remarkable comment, as
follows: “I think that only the Atomic Energy Society of Japan (AESJ) can describe thedelicacy of the issue of independency of regulation versus collaboration between
industries, academia, and the government. In the current opinion, people imagine thatdecision had been made based on the collusion between regulatory body, plant makers
and utilities. Therefore, people require perfect independence and purification fromindustries to regulatory body. In reality, it is impossible to perform regulatory inspection
for pumps if the design and operational condition of such pumps is not known. I also
think that real safety could never be achieved if the regulatory body does not understandhuman organization within nuclear power plants and the motivation of the workers
therein. The tendency to improve independency by purifying the organization has been
proposed as an example of a “no return rule,” but I do not believe that it will help secure
real safety.” 7-14
In reply to this comment the following remarks were added by Dr. Fuketa, aCommissioner of NRA: “The regulatory body must be independent, but it must not be
isolated. Moreover, people in regulatory body cannot secure the safety by themselves,
and they can just observe the actions to improve safety. These actions are performed by
the people in industry working at the facility. People in regulatory body must neverdiscuss about safety with neglecting the knowledge obtained from the field of realactions. Therefore, an exchange of opinions and the transfer of information should be
performed frequently. […] I believe that we are now at a stage of struggling to find agood system while public opinion is still critical. Transparency should be secured. But I
do not think that the current situation is good. The chances for good communication
with licensees are too few. The situation should be changed step by step. But I have tosay, “Please give us some time,” because we are carefully considering securingtransparency. Industries feel the need for communication, and we also feel the needmore deeply.” 7-15
The above issue has now already been clarified between the individuals
concerned and the NRA is expected to solve this problem in the near future.
Finally, the objectives of NISA are extracted from the first report by the Nuclear
and Industrial Safety Committee, which is the top advisory committee to NISA, as
follows: 1) To gain public trust; 2) To achieve effective regulation with scientificrationality; 3) To achieve crisis-management capability; 4) To provide a knowledge172basis; 5) and to implement human resources. 7-16
All of the above were previously not sufficiently achieved. Gaining the public’s
trust was considered to be the first priority, but NISA perfectly lost it. The objectives of
new regulatory authority, NRA, must be as same. NRA is expected to achieve the goalof NISA, in particular the goal of gaining the public’s trust.
(Drafter: Kiyoharu Abe)

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7.4

Issues on Nuclear Emergency Response
Nuclear emergency planning in Japan was established in relation to lessons
learned from the Three Mile Island Nuclear Power Station Unit 2 (TMI-2) Accident in
1979. In July of 1979, the Central Emergency Preparedness Council determined“immediate measures for nuclear power stations.” 7-17 In the case of emergencies
occurring in nuclear power plants, the following were developed and were deemed to be
the responsibility of the state: an emergency communication system between central andlocal governments, technical support organizational structures such as an emergency
technical advisory organization, emergency monitoring and an emergency medical
dispatch system. In June of 1980, the NSC declared “Emergency Preparedness andResponse Guidelines for the Vicinity of a Nuclear Power Station (Nuclear Emergency
Guideline)”. 7-18 Nuclear Emergency Guidelines represent the summary of technical andprofessional items, and are aimed to enable the implementation of smooth-runningemergency response activities in relation to specific nuclear emergency events.
After the Chernobyl accident occurred in the former Soviet Union in 1986, it
was emphasized that Japan’s reactor types were different from those of the former
Soviet Union. It was considered impossible that such an accident could occur in Japan,
but that accidents like TMI-2 were possible. As the result, lessons learned from
Chernobyl accident were slightly reflected by the preparedness of emergency response
system in Japan. The scope of an EPZ was discussed in the Nuclear Emergency
Guideline, but being sufficiently prepared for and developing a sufficient response to an
accident severe enough as containment vessel damage was not developed. 7-19 In
addition, long-term protective actions relating to temporary evacuation that occurred in
the Fukushima Daiichi accident were not developed.
After the southern Hyogo Prefecture earthquake in January 1995, responses tovarious types of disasters were determined and included in the Basic Plan for
Emergency Preparedness, in accordance with the Basic Act for Emergency Preparedness(Emergency Act). Immediately after the fire explosion accident in the asphalt
solidification processing facilities of the Power Reactor and Nuclear Fuel Development
Corporation Tokai reprocessing plant in June 1997, “Guide 10: Countermeasures against
a nuclear emergency” was added to the Basic Plan for Emergency Preparedness. This
Guide was discussed and developed by the Central Emergency Preparedness Council,
and the roles and responsibilities of the organizations involved in relation to nuclear
emergency preparedness and responses were clarified. In September 1999, a criticalaccident occurred in the Tokai-mura JCO uranium processing plant (the JCO accident).
Along with safety regulations for nuclear fuel facilities, various issues such as the rapid

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initial activities for emergency responses, cooperation between central and local
governments, strengthening of the national emergency response system and clarificationof utilities’ responsibility became apparent. In response to this, in December of 1999, asa special law of the Emergency Act and “The Act for the Regulation of Nuclear Source
Material, Nuclear Fuel Material and Reactors” (Reactor Regulation Act), a new law
called “The Special Act for Nuclear Emergency Preparedness and Response” (Nuclear
Emergency Act) was enacted and promulgated. The special parliament that enacted the
Nuclear Emergency Act, with cooperation of the relevant ministries and agencies,
summarized the measures that required budgeting for. These measures included: thestrengthening of communication systems, strengthening of radiation monitoring, theconstruction of an off-site center, the preparation of resources needed for an emergency
response, and the development of an emergency medical system.
As described above, the Emergency Act and the Nuclear Emergency Act are thetwo bases related to the nuclear emergency system in Japan, and in the Emergency
Preparedness Basic Plan the roles and responsibilities of related agencies are clarified.
The Nuclear Emergency Guideline is the guide for action and emergencies, and whennational government, local governments, and industries develop a nuclear emergency
preparedness plan, it is suggested that they follow these guidelines. Local emergency
plans were then developed based on the Emergency Preparedness Basic Plan and the
Nuclear Emergency Guidelines, in addition to utilizing the support of resources from
off-site centers.
The emergency response required for the Fukushima Daiichi accident wasmore serious than that of the JCO accident. However, the problems that arose such asconfusion of the initial response, insufficient cooperation of related organizations, andan unclear decision-making scheme, were similar to those experienced during the JCO
accident. The most important lessons are described as follows:(Lesson 1) As in accidents at TMI, Chernobyl, Goiania (a radiation source accident
occurring in Brazil), and JCO, the main reason for the failure of the emergency
response was that the operators and regulatory authority believed that such asevere emergency was impossible, and therefore the facility was ill-prepared.
Issues on radiation protection during the Fukushima Daiichi accident that werediscussed in the Seminar, such as the issues on protective action itself, issues on theemergency management, and the clarification of responsibilities and roles of relevant
agencies, are summarized.

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7.4.1 Issues on the Implementation of Emergency Protective Actions
The concept of operation in relation to various measures necessary forprotecting the health of residents during the emergency was not shared clearly. In the
Nuclear Emergency Guidelines, technical guidelines are described that refer to: therange of the emergency planning zone 7-20, the guide for implementation of theemergency response such as the criterion for notification of abnormal events andidentification of emergencies 7-21, protective actions 7-22, and indicators for protective
actions 7-23. However, no scheme or basic concept exists in relation to how thesemeasures should be implemented. The only sentence recorded states that “whenimplementing evacuation or sheltering inside houses, the scale of the abnormal eventand the weather conditions should be considered, and it is necessary to identify therange of the evacuation zone and implement evacuation in stages.” 7-24 In addition, when
the Nuclear Emergency Guidelines were reviewed in May, 2007 in light of theinternational experience of the IAEA, only this phrase was added: “Before or
immediately after the release of radioactive materials, based on the local situation andabnormal circumstances, preventive measures such as sheltering at houses andevacuation are effective”. 7-24 However, there was no clear guidance on which measures
should have the highest priority in terms of implementation, nor a detailed operational
concept in relation to these protective actions.
When responding to an emergency situation, it is important for relatedorganizations to share a decision-making scheme that is consistent with thedevelopment of the particular emergency situation over time. Figure 7.4-1 shows thestages of the emergency situation and the development of its management 7-25. It
includes three stages: preparation, response, and restoration. The response stage is
subdivided into an initial response and a medium-term response. As shown in Figure
7.4-1, the emergency exposure situation, which is described in the 2007
recommendations of the ICRP 7-26, should be applied to the initial and medium-term
response stages, and the existing exposure situation should be applied to the restoration
stage. The following relates to the issues concerning crisis management andconsequence management in the response to the Fukushima Daiichi accident.

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Response

Preparation

Planning
Stage

Early
Event/
InitialResponse

CrisisManagement

Intermediate
Transition to
RecoveryConsequence(including
Management
recoveryplanning)

Emergency Exposure Situation

Recovery
Late
Recovery/Long-term
Restoration
Existing Exposure
Situation

Fig. 7.4-1 Emergency management stages(1) Protection strategy and issues of preventive emergency protective action
Since the JCO accident, emergency response drills were frequently performed.
The Emergency Response Support System (ERSS), which predicts the progression of
the accident and the source term released from the plant, and the System for Predictionof Environmental Emergency Dose Information (SPEEDI), which conducts a predictionof exposure dose in the environment, are included in the emergency calculation
prediction system. Based on the prediction result, the scheme of evacuation andsheltering was determined. The projected dose obtained from this prediction system wasregarded as an indicator for implementation of protective actions (an intervention level).
For example, the effective doses initiating sheltering and evacuation were 10 mSv and50 mSv, respectively. In accordance with these statistics, ranges for sheltering andevacuation region were identified. As the result, people involved in the emergency
response tended to put their faith in prediction calculations. However, although thesystems themselves were working in Tokyo during the Fukushima Daiichi accident,
information concerning the reactor state could not be obtained because of the disruption
to the communication system caused by earthquake, and in the reports of Governmental
Investigation Committee 7-27, 7-28 concluded that the original function of the prediction
system could not be reflected in the emergency responses. 7-29, 7-30
At 20:50 on March 11 (6 h after the earthquake occurred), an evacuation
radius of 2 km was initially set by Fukushima Prefecture. Later, at 21:23, thegovernment altered the evacuation radius to within 3 km, and declared that those living
within a 3–10 km radius should shelter within their houses. At 05:44 on March 12, theevacuation radius was expanded to 10 km. Therefore, emergency protective actions
were implemented at an early stage of this accident, and such actions were based on the177loss of cooling function and the increase in pressure within the containment vessel in
Unit 1. At 18:25 on March 12, after the hydrogen explosion occurred in the reactor
building of Unit 1, on the basis of risk of a multi-unit simultaneous disaster civilian
evacuation was expanded to a range of 20 km. At 11:00 on March 15, after the eventsoccurred in Unit 2 and 4, civilians within a 20–30 km radius were instructed to shelter
within their houses. Because there was no preparation for implementing emergency
protective actions, the evacuation scope was not based on well-founded considerations.
However, as described in Clause 5.9.4, no civilians suffered by large exposure doses
which can cause deterministic health effects.
Without any source term information, results of SPEEDI’s analyses assuming aunit release of radionuclides were reported to the Nuclear Emergency Response
Headquarter (NERH) and NSC. However, this prediction result of SPEEDI was notavailable to the public, which caused social problems. Critics of this action claim that
non-disclosure of the SPEEDI results led to unnecessary exposure, because someresidents were evacuated from within the 20 km radius range on March 12 to thenorthwest of the Fukushima Daiichi site, which was then affected by serious pollution
on the afternoon of March 15. As illustrated in Clause 5.9.1, the changes in the weather
conditions in Fukushima Prefecture on March 15 were very complex. There was anortherly wind in the early morning, which changed to an easterly wind by noon. In theafternoon it had changed to a southeast wind, and by the evening there was rainfall fromthe north, and by late night the whole prefecture experienced rainfall. The prediction
result on March 15 by SPEEDI which was published, simulated accurately the
temporary movement of plume, but without the timing of the release and the change indirection of the release it was difficult to deliver accurate protective actions to theresidents living in the vicinity of the site, most of whom had already evacuated to thearea around 20 km from the site. There was no information of the dose level and theexposure route at the areas around 20 km from the site. Based only on the prediction
result of the plume’s diffusion direction, no recommendations for protective actions
could be made.
Based on a source term calculation with a severe accident analysis computercode MERCOR, which was submitted in the Report of Japanese Government to the
IAEA Ministerial Conference on Nuclear Safety in June 2011 7-31, 7-32, the deposition
patterns of Cs-137 were calculated by a Level 3 probabilistic safety assessment code,
OSCAAR developed by JAEA. Figure 7.4-2 shows this result and the monitoring result
by MEXT 7-33. Because MERCOR was not able to clearly analyze the amount of releasein Unit 2 on March 15, it was difficult to obtain simulation of the pollution distribution178

in a northwest direction. Even after the accident, as shown in the result, it was very
difficult to reproduce the exact source term after the accident. Even if exact source term
was given, it was still difficult to accurately evaluate the exposure dose at a distance of
10–20 km or more away from the site. Crisis management requires a quick decision in
the early stage of an accident where there is insufficient information present. In
consideration of the uncertainty of the source term evaluation and the environment
impact assessment, the scheme that depends on a prediction system requires changing.

Fukushima
Daiichi NPP

Fig. 7.4-2 Comparison of the calculated Cs-137 distribution using prediction codes
and the monitoring results

As revealed in the Chernobyl accident, one of the most important emergency
protective actions in the initial stages of crisis management is to take stable iodine as
this prevents the thyroid absorbing radioactive iodine. Therefore, in order to reduce thedose received by the thyroid, stable iodine must be taken as soon as possible, even after
exposure to a small amount of radioactive iodine. According to the Government
Accident Report, at 13:15 on March 12 the Local Nuclear Emergency Response
Headquarter (L-NERH) instructed Fukushima Prefecture and local governments
concerned to deliver stable iodine to shelters and to dispatch to pharmacists 7-34. In
addition, on the morning of March 13, the NSC replied to the L-NERH’s by sending theEmergency Response Center of NERH an instruction that stable iodine should be taken
if the screening result showed a measurement of more than 10,000 cpm 7-35. However,this instruction failed to reach the L-NERH 7-35. Therefore, although administration of179stable iodine to civilians at an early stage was actually considered and acted upon,residents living within 20 km who were evacuated were not given stable iodine. The
NSC recommended that patients in hospital within the evacuation area (20 km) shouldtake stable iodine on March 15, and on March 16 the same recommendation was madefor the remaining civilians within the evacuated area 7-36. However, again thisrecommendation was not implemented by the prefecture because there were notconsidered to be any remaining civilians sheltering within the evacuated area. In
contrast, in Miharu Town and in the vicinity of certain other local governments, stableiodine was delivered and taken based on their judgment. In Miharu Town, 95% of theresidents received stable iodine on March 15 by pharmacists 7-37.
The main reason for this confusion in administering preventive stable iodine isthat no clear operational concept existed for the implementation of protective actions. In
the Nuclear Emergency Guideline made by NSC, only a thyroid equivalent dose of 100
mSv for children was included in relation to indicators for preventative administrationof stable iodine 7-38. A proposal for the effective procedures for preventive taking was
described in “the administration of stable iodine during a nuclear emergency” which
was issued by the Emergency Response Technical Committee of NSC in April 2002 7-39,
but it had not been fully reflected in local emergency response planning. In particular,
despite the need to promptly administer iodine, the instruction should have come froman off-site center to NERH, then from the Chief of the NERH (the prime minister) to
L-NERH, then to the prefectural governors and residents. In this scheme, the final
decision was made in relation to the judgment of the national government and was notbased on a local decision. This scheme can easily prevent immediate decision making.
Confusion occurred on March 15 when the government decreed that civilians
in a radius of 20–30 km from the site should shelter in their houses. It was difficult topredict when to lift the order of sheltering in houses because the plant's condition was
unstable, but the lack of the infrastructures made the long term sheltering difficult. On
March 25, therefore, voluntary evacuation was decreed.
It is common in the procedures of local emergency guidelines to orderevacuation in areas with a high exposure level, and sheltering is recommended for thoseliving within a certain range from the high exposure area. The Fukushima Daiichi
accident has revealed that it was very difficult to decide who should shelter within theirhouses, and who should be evacuated. It is evident that remaining indoors is quick toimplement and that civilians who are sheltering are able to easily access new
instructions. However, any reduction from high exposure is dependent on the structure180

of the building. Long term sheltering was not realistic. This measure is temporarily
effective as it reduces the initial exposure to radioactive materials, but evacuation isthen essential. It is necessary to carefully review in advance implementation proceduresof such protective actions, including sheltering, evacuation, temporary relocation, andadministration of stable iodine.
As described above, in the Fukushima Daiichi accident, urgent protective
actions were carried out in a way that was different to that decreed by the guidelines. In
the process of communicating with residents, implementing evacuation beyond EPZ,
and sheltering, there was considerable confusion. Fortunately, as already mentioned in
Clause 5.9.4, civilians did not receive significant exposure exceeding the threshold
amount. However, due to poor transport in harsh conditions, many lives were lost in theevacuation of patients from hospitals and nursing homes within 20 km. Therefore, it isevident that during emergency conditions special consideration should be applied to theprocess of evacuating hospitals and prisons. In the report of “the uranium processing
plant’s critical accident investigation committee” in Tokai-mura 7-40, it had been
mentioned that special protective actions should be considered for vulnerable people 7-41,
but again this was not realized and provided for.
(Lesson 2) Emergency protective actions should be performed using criterion
prepared in advance in relation to the condition of the facility. Emergency
protective actions should not be performed based on a prediction of propagation of
the incident and/or a radiation dose prediction. Precautionary protective actions in
an area (that have been decided in advance) should be performed quickly before
radioactive materials are released.
(Lesson 3) Evacuation and transfer should be implemented safely. In other words,
there should be no life-threatening events occurring during the ongoing evacuation.
Sheltering in houses should be carried out only in the short-term period before theevacuation and transfer is considered possible.

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(2)Protection strategies and issues; and consuming food and drink
Large amounts of radioactive materials were released into the atmosphere,
mainly between midnight on March 14 to early morning on March 16. This caused
contamination of certain foods and drinks. The ingesting of contaminated substanceleads internal exposure, and the excessive fear against the internal exposure may cause
the harmful rumors concerning the products and people which may affected by thecontamination. Therefore, the need for a reasonable protective strategy in relation tocontamination of foods and drinks is essential. Two main issues have been raised inrelation to the intake restriction and the trade limitations on contaminated substances.
Firstly, in consideration of the contribution of the short half-life of radioactive
nuclides such as iodine and tellurium, prompt trade and ingestion limitations shouldhave been implemented soon after the plume had passed.
On March 16, iodine-131 exceeding the intake limitation was detected inwater, milk, and leafy vegetables within an area outside the 20 km evacuation zone. As
mentioned in Clause 5.9.3, the Ministry of Health, Labor and Welfare (MHLW) who
hold the jurisdiction on food safety had not previously established any limits oningestion related to radioactivity. On March 17, a provisional limit for food and drink
intake was decreed. Figure 7.4-3 shows changes in the concentration of iodine-131 in
tap water in Fukushima Prefecture, Ibaraki Prefecture, and Tokyo. Radioactive materials
including iodine-131 that were released into the atmosphere on March 15 fell on the
ground as a dry deposition, and as a wet deposition within rain, which caused them tothen flow into the river and the raw water aqueduct within a short period of time; theeffective half-life of iodine-131 in water was observed as being about 2.8 days 7-42.
Head of Water Supply Division, Health Service Beau of the MHLW decreed a“response for water due to the Fukushima Daiichi and Daini nuclear accidents,” 7-43 on
March 19, and a “tap water intake by infants” 7-44 on March 21. Water suppliers were
instructed to announce to citizens the need to refrain from drinking tap water in excessof the provisional regulation value. As shown in Table 7.4-1, Kinase et al. reported thatthe highest averted thyroid equivalent dose to infants by the restriction of tap water
intake was estimated at about 8 mSv in Iitate Village 7-42. As shown in Figure 7.4-3, theamount of iodine-131 in tap water did not uniformly change, reflecting regional
characteristics such as pollution condition, weather, and the water intake environment.
The possibility of ingesting iodine-131 was considered likely during a period of fivedays from March 16 when it was first detected to March 21 when the first restriction
was issued.

182Prompt action against substance which can be ingested by people directly isessential to avoid unnecessary exposure. It is of note that the quick judgment taken on
the intake limitation was determined by an increase of the air dose rate due to thepassage of radioactive plume and its deposition, not by the indicators of radionuclide
concentration that needed time-consuming measurements.
In this accident, 20 water suppliers in Fukushima, Ibaraki, Chiba, and Tochigi
prefectures and Tokyo were involved in formulating the intake limit for infants, and thewater utility of Iitate Village in Fukushima Prefecture decided the intake limit for
general civilians. The water intake limit began on March 21 and was lifted by the Iitate
water utility on May 10 for infants, and on April 1 for general civilians. For most ofother water utilities, the limit lasted for two or three days, and nine days at most.
According to the analysis of Kinase et al., 7-42 the effective decay half-life of iodine-131
concentration in water is approximately 2.8 days, and it tends to decrease relatively fast.
Radioactive cesium did not reach the provisional regulation value (200 Bq/kg) because
it was absorbed by the soil, although the part that reached water treatment processes via
rain was removed in processes such as coagulation sedimentation 7-45.

Fig.7.4-3 I-131 Changes in iodine-131 concentration within tap waters

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Table 7.4-1 Water intake restriction and thyroid equivalent dose value

Administration Area

Began

LiftedIitate Village, Fukushima

March 21

May 10Averted thyroid
equivalent doseto infants (mSv)
8.3

Koriyama City, Fukushima
Kawamata Town, Fukushima
Minamisoma City, Fukushima
Iwaki City, Fukushima

March 22
March 22
March 22
March 23

March 25
March 25
March 30
March 31

0.51
1.1
1.7
2.9

Tokai Village, Ibaraki
Shinjuku, Tokyo

March 23
March 23

March 26
March 242.1
0.13

The delay in implementing restrictive measures on ingestion of food and drink
by the former Soviet Union government was the primary factor involved in the increaseof thyroid cancer in children after the Chernobyl accident, and it was considered thatmilk contaminated by iodine-131 was the main exposure pathway. As illustrated inclause 5.9.3, in the Fukushima Daiichi accident on March 16 a very high concentrationof iodine-131 of up to 151 kBq/kg was detected in soil, up to 1440 kBq/kg was detected
in weeds, and iodine-131 that was found in raw milk and leafy vegetables (spinach),which exceeded the provisional regulation value. On March 17, the NSC advised the
NERH to forbid the intake of products, but it was not until March 21 that the MHLW
instructed restrictions for trade measures. There was also a delayed judgment incarrying out protective actions in relation to the ingestion of tap water, and no clear
explanation for such a delay has since been provided. The restriction on the intake offood and drink was an effective means of reducing internal exposure, but it wasconsidered that social concern in relation to rumors was probably one of the reasons forthe delay of implementing such restrictions. The decision as to what the size of the area
should be in which protective actions were taken was also difficult to achieve. Local
government firstly set a restriction over a wide area; then, in accordance with the basisof the “consideration of the inspection plan, restricted items, set-up area, and release”
published by NERH on April 4 7-46, local government set reduced sections of theiradministrative area, and hoped to avoid the impact of rumors.
Concentration values stated in the Nuclear Emergency Guideline by the NSC
are indicators that the NERH should start considering about the restriction of ingesting
if the concentration of radionuclides exceeded the value 7-47. In order to avoid184

unnecessary exposure and to speedily determine these restrictions, however, a two-stage
scheme is required. In the first stage, an operational intervention level (OIL) for the air
dose rate should be used to restrict the intake of contaminated food temporarily, as
proposed in general safety guide by IAEA 7-48. Later, in the second stage an OIL basedon the concentration of radioactive materials should be used to restrict trade and intake.
Another issue relating to restrictions on shipping and the intake of food anddrink is how to set the restriction level to protect from long term effects of radioactive
materials such as cesium. In the Chernobyl accident, the Soviet Ministry of Healthintroduced temporary permissible levels (TPL) for radionuclides in food and drink. In
terms of TPL in 1986 (in the first year after the accident) a level of 50 mSv was
assigned as an internal exposure limit, which was half of the whole body dose
restriction of 100 mSv. In 1988, the TPL reduced this to 8 mSv, and in 1991 to 5 mSv.
With regard to the policy of the former Soviet Union (later three countries), based on
the improvement of the environment due to the decay of radionuclides and penetrationand immobilization of radionuclides in soil, criteria for radiation effects and TPL was
reduced, and hence the restriction on the content of radionuclides for food producers
was lower and stricter. The value of the TPL was established by taking into account thebalance between the need for reducing internal exposure and the need of maintaining
agriculture production in the restricted area.
After establishing a provisional regulation value in the Fukushima accident, inthe light of a report by the Food Safety Commission dated October 27 7-49, 2011,assessing the health impact, MHLW conducted the discussion at a working group of thePharmaceutical Affairs and Food Sanitation Council, and set a new standard value froma long-term perspective, and this came into force on April 1, 2012. The regulatory
indicators in food and drink, as a derived intervention level (DIL) in general, arecalculated as follows: DIL = RL / ( f × I ×DF ); where DIL: derived intervention
level (Bq/kg), RL: reference level (mSv), f: pollution rate, I: intake amount (kg), and
DF: dose conversion factor of intake by mouth (mSv/Bq). Since the intervention doselevel RL for food intake was reduced to 1 mSv from 5 mSv per year, the provisional
regulation value of 500 Bq/kg for radioactive cesium in vegetables, cereals, meat, eggs,
fish and others was reduced to one fifth (100 Bq/kg) as a new standard. The MHLW
explained that the provisional regulation value was able to secure safety, but from theperspective of further ensuring food safety, the reference value was reduced. Thisdecision was based on following reasons: (1) the codex commission uses 1 mSv per
year as an international standard for food 7-50; (2) according to the current monitoring
result, the concentration detected from food has shown a considerable downward trend185

over time. The concept of the assessment by the Food Safety Commission (which the
new standard is based on) is only related to the effects of intake on human health anddoes not change regardless of whether there is an emergency situation or normal
conditions. However, the purpose of the standard of the codex commission that wasreferred to in the MHLW’s explanation is different, and this is the international standard
relating to the import and export of food after nuclear accidents. It appears however thatreason (2) above encourages the efforts of producers to further reduce the nuclide
content; this occurred after the Chernobyl accident.
In accordance with the recommendation of the ICRP 2007 7-26, a single
protective action (such as a restriction of the intake of food and drink) should be
performed in the process of optimizing overall protective actions. Careful consideration
listed as follows is therefore required:- The avoidance of radiation effects by restrictions of intake and the securingof replacements with consideration of nutrition- The effect of an oral ingestion dose should be taken into account whensetting the dose-level standard-

A realistic assessment of eating habits (intake amount I) and pollution rate f
Food safety for consumers and consideration for producers in the affected
areas- Working in accordance with international standards
In the codex international standards, the pollution rate is set at 0.1. Though MHLW’s
new standard adopted the concept of general food safety regulation, pollution rate usedin the new standard is 0.5, which was adopted by the guideline of former NSC from theviewpoint of post-accident intervention. Therefore, the derived intervention level isextremely strict in Japan compared in comparison with international standards. Foodsafety is a great concern for the public in daily life. In addition to the trade-off between
consumers and producers, consumers who require zero risk and distributers who reply
to this requirement further complicate the problem.
(Lesson 4) In the restrictions on food and drink at the initial response stage ofcrisis management, an operational intervention level (OIL) that refers to data of
air dose rates should be used.
(Lesson 5) Realistic recommendations are required for protective actions inrelation to food and drink on a long-term basis, taking into account the situation in
affected areas and in agreement with international standards

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(3) Lifting of emergency protective actions and the implementation of long-term
protective actions
As described in Clause 5.9.4, on April 22 the NERH set up a formal plan for
the Deliberate Evacuation Area and Evacuation-Prepared Area. NERH later lifted
emergency protective actions and transferred to long-term protective actions, as
illustrated in Clause 5.9.6. At the time the accident occurred there was no indicator for
long-term protective actions in the Nuclear Emergency Guidelines, such as temporary
relocation. In 2006 before the Fukushima Daiichi accident, there was some discussion
on perspective issues during a meeting of the Nuclear Emergency Guideline working
group. The conclusion was recorded as, “in terms of long-term protective actions, and incomparison with short-term protective actions, it should be noted that the judgment of
implementation is very complex as many options exist within the different relatedorganizations.” 7-51
In 2011 when the accident happened, the actions were carried out
internationally in accordance with ICRP 2007recommendations 7-26 that had evolved
from the previous process-based protection approach, using practices and interventions
by moving to an approach based on the exposure situation and the recognized planned,
emergency, and existing exposure situations. In 2009, Publication 109, “application of
the commission's recommendations for the protection of people in emergency exposure
situations,” 7-52 and Publication111, “application of the commission's recommendations
to the protection of people living in long-term contaminated areas after a nuclear
accident or a radiation emergency,” 7-53 were published. On the basis of these, on March
21 2011 the revised basis safety standards (BSS) 5.0 was drafted by the IAEA andapproved by the Safety Standards Committee (CSS) in May. Instead of evaluating theeffectiveness of individual protective actions, a series of protective actions considered tobe most appropriate were created for use in the overall protective strategy by taking into
account all of the exposure pathways and all the protective actions. This approach aims
to be the best response overall in a situation where one individual protective action is
not enough. However, at the time of the Fukushima Daiichi accident, the Nuclear
Emergency Guideline was based on the previous ICRP 1990 7-54 recommendations andthe BSS of the IAEA before revision 7-55. Based on the 50 mSv of avertable dose applied
to evacuation, therefore, the NSC started consideration about the necessity of additional
evacuation.
Finally on April 10, the NSC considered the message from the ICRP 7-56 andadvised the implementation of a temporary relocation, which is called deliberate187evacuation. The instruction for sheltering was lifted and the area was set as an
Evacuation-Prepared Area. The range of the Deliberate Evacuation Area was based on
the concept of the ICRP’s 2007 recommendations, and 20 mSv per year from thereference level of 20–100 mSv of a residual projective dose was selected for the
optimization of protective actions. If the area was beyond 20 mSv per year, the residents
living there would be requested to temporarily relocate. In addition to the lack ofadvanced consideration for long-term protective actions, it took a long time to reach asufficient consensus because the international concept on radiation protection in
emergency condition had just changed before the accident, and this caused a delay inapplying the additional protective actions. Furthermore, it took time to communicate
with local government because some residents were forced to temporarily relocate. As
described in Clause 5.9.5, according to the health investigation by Fukushima
Prefecture the residents in the Deliberate Evacuation Area received a relatively highexternal exposure cumulative effective dose, due to delays in decision-making. As
mentioned in Clause 5.9.4, IAEA modified the OIL of evacuation that was used in the
Fukushima Daiichi accident and advised the Japanese government to carefully assess
the situation on March 30. These facts also show the importance of a judgment scheme
for implementing emergency protective actions after the release of radioactive material
to the environment, which includes both monitoring and an OIL based on the measured
amount in the environment.
(Lesson 6) The concept and judgment criteria involved in the implementation of
emergency and long-term protective actions, and the return to normal life, shouldbe established in the preparatory stage of an emergency. For protective actions
corresponding to an anticipated emergency, the guidance on applying radiation
protection principles should be pre-established.
(Lesson 7) The operational intervention level (OIL) as a guide for decision-making
during an emergency is regarded as being very important. More detailed
international guidance is needed.
7.4.2 Issues on Crisis Management and Operation of Emergency Response
Off-site emergency response is placed on the fifth level of the DiD. In IAEA's
fundamental safety principles 7-57 however, “emergency preparedness and response” isstipulated in Principle 9, and is separated from the primary means of preventing andmitigating the consequences of accidents that stipulated in Principle 8.The primary188

goals of emergency preparedness and response are: (1) To ensure that arrangements arein place for an effective response to a nuclear or radiation emergency at the scene, andas appropriate on a local, regional, national, and international level; (2) To ensure thatfor reasonably foreseeable incidents, any radiation risks would be minor; (3) If any
incidents do occur, to take practical measures in mitigating any consequences for humanlife and health, and the environment.
Regarding primary goal (3), the issues concerning protective actions in the
Fukushima Daiichi accident are analyzed in Clause 7.4.1. In relation to primary goal (2),in IAEA general safety requirement GS-R-2 7-58, the full scale of postulated eventsshould be considered in the threat assessment. Emergencies involving a combination ofa nuclear or radiological emergency and a conventional emergency (such as anearthquake) should be considered 7-59. Emergency plans in Japan were prepared with theassumption that the integrity of the containment vessel was ensured, and the necessityof emergency procedures such as evacuation is small enough at the case. Preparations
for combined disasters including earthquakes were not considered, in spite of lessons
learned from the Niigata Chuetsu-Oki earthquake. According to these, emergency
responses were insufficient in all nuclear utilities, and by local and national government.
No matter what accident occurred, there would have been a poor response
resulting from not being prepared for an event that was beyond the hypothesized range.
Therefore, in an assessment of threat, the ability to prepare for reasonably foreseeable
events is important, although in addition a flexible response is needed to events that are
beyond our ability to presently hypothesize. In order to ensure a potential emergency
that is within the hypothesized range, in the preparation stage efforts to broaden thepossible range of responses should be made in times of peace. To deal with the crisis
management stage, predetermined measures should firstly be established. In addition tothis, the capability to achieve a flexible response is also important for any situation thatis beyond expected. A clear allocation of roles and responsibilities among all responseorganizations at a local, regional, national, and international level is required, andreviewing any method involved in adjusting to allocation should receive feedbacks fromvarious training providers.
The NSC has been in discussions related to compiling Review Guide for
Nuclear Reactor Siting, and discussions on the relation between site evaluation, accident
management, and the emergency response have occurred. The document“Systematization of Safety Review Guides” published in 2003 stated that a “nuclear
emergency plan is set for national and local government to ensure that disasters areprevented, or to reduce the effects of radiation so that they are as low as possible.

189

Prevention of disaster should be secured by the technical framework of DiD and theseparation of nuclear facility from the public. Being prepared for emergencies is locatedon the outside of the technical framework of the DiD and the separation from the public,
and is a part of DiD in a broad sense. Therefore, emergency response plan is anadministrative measure which is independent from safety regulation based on Reactor
Regulation Act. Regulatory body should not consider the emergency response plan as
the requirement for site evaluation for reactor establishment permit.” 7-60 In addition,
this document defined that the accident management is an utility's voluntary action toenhance the safety operation. It means that the accident management was not the targetof regulatory requirement based on the Reactor Regulation Act.
In Japan, an Emergency Response Plan is based on the Nuclear Emergency Act,and is not a scope of the Reactor Regulation Act. There is a clear distinction in thatgovernments have jurisdiction for off-site activities and nuclear utility has a
responsibility for on-site activities. In addition, reflecting from the lessons of a lack ofinitial response in the JCO accident, within the Nuclear Emergency Act the national
government was assigned the dominant role of dealing with the intensive managementof emergencies. The allocation of roles and responsibilities is therefore inconsistent withthe consideration of emergency management shown in Figure 7.4-1. In the Fukushima
Daiichi accident it took 2 h and 18 min for the NERH to declare a nuclear emergency
situation in accordance with Article 15 of the Nuclear Emergency Act after TEPCO
reported the corresponding events to NERH. It then took 2 h and 20 min to declare thefirst instructions to evacuate the residents from the declaration of nuclear emergency
situation. In the initial crisis management stage where there are large uncertainties andminimal information provided, it is necessary to establish a scheme whereby local
government rapidly implement a predetermined emergency protective action incooperation with the nuclear utility that are closest to the disaster site. To achieve this,as clarified in the new Guides for Nuclear Emergency Preparedness issued by NRA,
many measures need to be considered, including the need for protective actions to becommonly recognized by relevant people, the acknowledgement of a need for acategory of emergency, the EAL to then judge the required category, and finally theannouncement of the emergency. Some tasks, such as the judgment of emergency
protective actions to the residents and its recommendation, are strongly related with theRegional Plan for Disaster Prevention, which is applied to natural disasters.
The special characteristics of a nuclear emergency are often emphasized.
However, the actual protective actions taken, such as evacuation and sheltering relate toother natural disasters, and only the range and form of such measures are different.

190

Local government is in charge of the operation of such measures, and the police, firebrigade, and the Self-Defense Forces stand at the forefront. In this sense, the operational
implementation of an emergency protective response to a nuclear emergency should be
integrated with the response to other more common disasters. It is explained that theoff-site center located at 5 km from the accident site could not play its original role
during the Fukushima Daiichi accident, because of the lack of robustness against natural
disaster. The failure of emergency power sources and the loss of communication tools
could be a part of the reason. Even if these infrastructures were available, however, theefficiency of off-site center is doubtful because it was not regularly used. The assemblyand delivery of human resources from outside during an emergency can become easily
confused. Considering an integrated approach between a nuclear emergency and natural
disasters, facilities and personnel within an emergency center (as established in every
prefecture) should be used effectively.
(Lesson 8) An emergency response should take into account all the possible
hypothesized events, including events with an extremely low probability ofoccurrence, and should be ready for a combination of emergencies such as thosethat come with earthquakes.
7.4.3 Conclusions
The implementation of a nuclear emergency response is intended to minimize
disastrous consequences for people, property, and the environment in the event of any
nuclear or radiological emergency. To achieve this goal, the requirements are specified
in IAEA safety requirement No. GS-R-2 as follows: “The response to a nuclear orradiological emergency may involve many organizations. The functions of many ofthese organizations would be the same for a nuclear or radiological emergency as for aconventional emergency. However, the response to a nuclear or radiological emergency
may also involve highly specialized agencies and technical experts. Therefore, in order
to be effective, the response to a nuclear or radiological emergency must be wellco-ordinated and arrangements must be appropriately integrated with those for aconventional emergency. In addition, the many misconceptions prevalent concerning
nuclear and radiological emergencies and the possible health effects of radiation
exposure could lead to inappropriate actions being taken. Consequently, preplanning on
the basis of established principles of radiation protection and safety is essential. Such
preplanning can be achieved only through a co-ordinated approach.” 7-61 (underlining
courtesy of the author).

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Hence, requirements for: the clear allocation of responsibilities among all
response organizations; well defined agreements between these organizations; andarrangements for coordinating an integrated response, must be implemented andreviewed by training.
(Drafter: Toshimitsu Homma)

192

7.5
7.5.1

Issues on Cooperation and Communication during the Accident
Outline of the Problem
During and soon after the Fukushima Daiichi accident, important information
was not shared among relevant organizations, particularly between the site and theregulatory body/government.
Even when some deficiencies may appear regarding safety design or daily
management of nuclear facilities, or some events may occur at those facilities, the majorrole of regulatory body is to provide oversight, while those problems are in the secondor third level of DiD. If an event progresses to severe accident conditions, however,regulatory body, country, and local governments should have their own roles. Under
such situations, all organizations must know what is happening and understand what
measures other organizations are implementing in turn.
However, cooperation between organizations was insufficient during the
Fukushima Daiichi accident. In addition, plant parameters, which are indispensable forunderstanding what happened, and information regarding their existence and credibility,
were not communicated timely and efficiently.
We (AESJ-NSD) considered this problem to be one of the most importantduring our investigation, although we had neither the authority nor time for further
investigations. Here we list some issues related to this problem, although we recognizethat this analysis is not exclusive. We hope that the NRA will reconfirm the issues andadopt appropriate responses.
7.5.2

Organizational Roles when Severe Accidents Occur
When a serious accident, e.g. a core melt accident occurs, cooperation between
regulatory body, country, local governments, and the like is essential in addition tooperator’s actions. The role of each organization was introduced by Mr. Minoru Saito
from the Japan Nuclear Energy Safety Organization (JNES) at the fifth Seminar 7-62.
The role the off-site center played during the nuclear accident caused by the earthquake
and tsunami was also discussed, and the issues to be addressed were defined. Mr. Saito’spresentation can be summarized as follows:(1)

Initial action at the off-site center during the Fukushima Daiichi accident
Local nuclear emergency response headquarters (L-NERH) were established at
the off-site center under nuclear emergency situation. Its major activities includes: i)information sharing among relevant organizations and ii) the coordination of response
efforts. The off-site center was situated 5 km from Fukushima Daiichi Nuclear Power

193

Station, and the spare facility is prepared in Minamisoma city. Its function was
relocated to the Fukushima Prefectural office, local government, in Fukushima City at
the later stage of the accident. In addition, Japan’s Air Self-Defense Force’s
Ohtakineyama Sub Base was used for human resource transfers from Tokyo.
At 10:30 on March 12, 2011, initial action at the off-site center was performed.
The objective of an early date at the off-site center was to collect information about thestatus of the implementation of the evacuation order as well as the conformation ofplaces and number of iodine tablets were supposed to be prepared. The Off-site center
activities were changed gradually. The triggers for this change were radioactive releases
into the environment and the change in the plant accident conditions. Liaisons included
the following:First, determining whether radioactive release occurred was important, as therole of the off-site center depends upon this factor. If radioactive release occurs,
screening and decontamination are required.
Second, the coordination of response activities is required if some large eventsoccur on that site. For example at Fukushima Daiichi, the off-site center played
miscellaneous roles regarding to on-site as well as off-site after the hydrogen explosion
on March 14, 2011. Since there was no suitable place for discussing and conducting
such activities except for the off-site center, it became the place for planning theshipping of fire engines for water injection and for receiving the operators who were
injured by the explosion While the original role of the off-site center was assumed to be
information sharing and the coordination of activities, in fact it also became the place
for real operations like a first-aid post.
In parallel with the activities mentioned above, the off-site center was involvedin activities to share the information on the public evacuation status. Through these
activities, issues were revealed on the infrastructure for information sharing under thecomplex disasters. The full-scale nuclear emergency preparedness network was
unavailable because optical cables between relay centers were broken. Communication
between the off-site center and the ERC was performed using satellite phones andfacsimile. Securing electrical power was especially difficult because emergency diesel
power generators failed due to seismic damage to the fuel supply system, although thisfailure was repaired by March 12, 2011. On March 15, the air dose rate at the off-site
center increased to 100 Sv / h, and the alarm was sounded intermittently. Therefore, at
9:26 on March 15 it was proposed to the Minister of the Ministry of Economy, Trade,and Industry to move the off-site center to the prefectural office after conforming194evacuation accomplished. This demonstrates that requirements for the location andfunction of off-site centers should be improved.
The followings are the lessons regarding to the activities of the off-site center:
Face-to-face confirmation is necessary in order to implement adequate protective
measures for residents. Plans for responding large radioactive releases at an early stage
are important, and initial action for protecting residents should be planned and carriedout accordingly. The formulation of response schemes should be required for complex
disasters and nuclear emergencies initiated by external initiators. Clear assignment of
the chain of command and roles of involved organizations is necessary for information
sharing among local relevant people and to coordinate actions.
(2)
Discussion by the United States and international society regarding theissues highlighted by the Fukushima Daiichi accident
The issues that arose from the Fukushima Daiichi accident prompted
discussion in abroad on response schemes at early stages and the long term
countermeasures.
The response to high amounts of radioactive release at an early stage is based
upon the Emergency Action Level (EAL) and Operational Intervention Level (OIL).
The distance from a nuclear facility is decided in advance using the EAL threat categoryof the facility, which is also used to determine whether precautionary evacuation fromthe area should be taken. If radioactive release occurs, decisions are made according to
the OIL and monitoring results. Gradual evacuation based on the OIL after radioactive
release may lead to some exposure, although this risk can be avoided by precautionary
evacuation included in staged-evacuation based on the EAL.
If the plant condition is harmful and the situation cannot be rectified after
evacuation of the precautionary action zone (PAZ) is performed, evacuation in the nextarea would be considered. The question “how long will it take to finish the evacuationof residents in PAZ?” is important in this consideration. In other words, the Evacuation
Time Estimation (ETE) is necessary when examining the validity of the evacuation plan.
If residents living outside of the PAZ evacuate voluntarily using private automobiles,
evacuation inside the PAZ would take longer. Therefore, management of residents
around the PAZ is important when evacuating residents inside of it. The ETE will vary
according to the population around the nuclear facility and the shadow-evacuation ratio.
These factors will affect the effectiveness of reducing exposure risk and develop theplan for evacuation.

195

(3) Discussion following Mr. Saito’s Presentation 7-63
In the discussion after Mr. Saito’s presentation at the fifth Seminar, aparticipant mentioned that “there are many stakeholders, and accident preparedness
should ideally be discussed with country and local governments.” In contrast, anotherparticipant mentioned that “outside support to the site should be limited to what iswanted by the utility, as unnecessary intervention will add confusion. It is important todecide beforehand what should not be done by the outside people.” Relating to thiscomment, another participant mentioned that "if too many people try to act negative
results may occur, even with the best of intentions. A system for gathering staff isnecessary, who should be selected according to their assigned roles."Mr. Kosaku, a staff of NRA, explained the current situation as follows:“The consideration of emergency response systems has been in progress since the NRA
was established. The operator will prepare an emergency response center during an
accident, where NRA staff will also be present to participate in the discussion. As aresult, requests from the operator will be comprehended quickly. Information will besent directly from the center to the Cabinet and Prime Minister. The Cabinet will
coordinate the communication and collaboration of activities between other ministries
such as the SDF to ensure that response systems are improved. The Nuclear Emergency
Council will establish a collaboration system between ministries in preparation of these
activities beforehand. The effectiveness of these actions cannot be properly evaluated asof yet because they have just been implemented, but we would like you to be aware of
the situation regarding accident preparedness.”
Dr. Fuketa, Commissioner of NRA, concluded the discussion with thefollowing:“It is easy to imagine that answering variety of questions would be a heavy burden
under so confused situations due to serious accident. On one hand it is important not to
obstruct operators’ activities during an emergency but on the other hand the government
should respond as much as possible to the requests from the operator.”
7.5.3

Issues on Information Sharing during the Fukushima Daiichi Accident
The plant parameters of Units 1 and 2 could not be obtained for a period oftime because their DC power was lost by tsunami. The batteries in Unit 3 were notsubmerged by tsunami but DC power was later lost when these batteries were exhausted.
The operators attempted to get plant parameters by connecting the available batteries byhand. Investigations afterwards showed that there is no credibility in the reactor waterlevel measurement during severe accident due to instrumentation mechanism.

196

According to the situation described above, it was difficult even for the operators on-site
to comprehend the accident situation correctly. Furthermore, the information about such
instrumentation problems was not communicated accurately and timely to theregulatory side organizations and national and local governments. These issues are
discussed in this clause.
(1)

Credibility of the reactor water level indication
It is already well known that the Fukushima Daiichi accident is categorized as a“loss of all AC power accident” or a “station blackout accident” from the viewpoint of
event categorization which is used in PRA. Furthermore, the accident was also followed
by a “loss of the ultimate heat sink” and in some units by a “loss of DC power”.
However, from the viewpoint of thermal hydraulics or accident progression, this
accident is a very typical “coolant boil-off accident” where the reactor core is heated up
according to boil-off of the reactor coolant due to loss of heat removal capacity.
In a coolant boil-off accident, the reactor water level is the most influential
parameter onto the accident progression. If the water level is maintained higher thanabout 55% of the Active Fuel length, which is almost equal to TAF-1,600 mm, thecladding temperature does not reach an extremely high level, intense zirconium-water
reaction rarely occurs, and hence neither a high amount of generated hydrogen nor asevere accident is likely. If the water level decreases below this level, however, thezirconium-water reaction will be severe. This reaction is exothermal and has a positive
feedback. Therefore, high amounts of hydrogen will be generated and a core melt will
occur soon after the water level drops below this level.
On the afternoon of March 12, 2011, TEPCO reported to the regulatory bodyand other relevant organizations that the measured reactor water level of Unit 1 wasstable between TAF-1800 mm and TAF-1600 mm. This information indicated that water
injection was somehow achieved and the amount of injected water was balanced withthe boil-off caused by decay heat. Under such situations, however, hydrogen explosion
occurred at 15:36. It revealed that the water level indication by that time was incorrect.
It took two months for TEPCO to announce that the measured water level had
been incorrect. On May 12, 2011, they reported that the reactor water level in Unit 1
was clearly lower than the previously published value. The reason for this mistake was
explained to be due to dry out of the reference rig.
The water level values were informed to relevant people just as “Measured
Reactor Water Level” throughout the accident. It might mislead people to understandthat (a) DC power had no problem (measurement was being done!) and (b) the reactor

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core was still cooled by water or steam, causing the accident conditions to beinaccurately assessed.
(2)

Information on the loss of DC power and the failsafe design of IC
Many of the personnel, who were involved in the accident response activities
but far from the Fukushima Daiichi site, incorrectly assumed throughout the accident
that “DC power was alive but would possibly be lost within a time frame of aroundeight hours.” It took a long time for them to recognize that the DC power had alreadybeen lost because the batteries were submerged by the flooding.
We presume that April 4, 2011 was the first time that the regulatory bodyofficially explained this DC power loss. This date was the first day of the Fifth Review
Meeting at the Convention on Nuclear Safety (CNS) held at IAEA in Vienna, where the
Deputy Director-General of NISA addressed the Fukushima Daiichi accident at the“Side Event” of CNS. Although the IAEA member states who did not sign the CNS
treaty cannot join the CNS Review Meetings, the Side Event was open even to thosecountries. He explained that the DC power was lost by water and plant parameters were
measured using batteries brought in from outside. Although we did not perform further
investigation, NISA may not have recognized the loss of batteries until that time.
It was also incorrectly understood that the loss of DC power occurred in all 1 to
3 Units at that time. In fact, however, DC power in Unit 3 was not lost at the time offlooding. This information was announced at the IAEA’s mission at the end of May,2011.
If information such as the aforementioned is not shared by relevantorganizations, accident analyses and corresponding measures will be incorrect.
The next concern after we knew the loss of DC power was whether core
injection was possible without DC power. If we look at the speed of accident
progression at Unit 1 and Unit 2, it suggested that the isolation condenser (IC) in Unit 1
had almost lost function resulting in early core melt and that the reactor core isolation
cooling system (RCIC) in Unit 2 remained functional very long time. However, thequestion as to why this happened remained.
It took a very long time for the announcement that the IC in Unit 1 wasdesigned to be “fail-close.” Accident outcomes are almost the same under the following
two conditions: 1) IC line lost its function because the line was closed when the tsunami
struck; and 2) “even if the IC line was opened, it would be closed by the fail-closesetting at the loss of DC power. However, this resulted in significant differences in198

analyses of accident causes. Thus, such important information should be shared quickly.
The issue on continuous cooling by RCIC at Unit 2 is discussed in Item (4).
(3)

Plant parameters before the tsunami struck
Relevant people and organizations besides TEPCO believed for a time that theplant parameters were impossible to obtain because DC power was lost in the tsunami.
It was much later when it was deduced that “there must be records of plant conditions
measured using normal processes before the tsunami struck because DC power was
active by that time, enabling isolation valves to display their condition and becontrolled.”
With regards to this, NISA later received an explanation from TEPCO that it
was hard to collect these records from the site under such high radioactive levels.
However, TEPCO should have voluntarily disclosed that such important data
existed.
(4)

Credibility of measured reactor water level at Unit 2
The RCIC in Unit 2 continued operation with its open isolation valves. The

reactor water level in Unit 2 was maintained high enough for a long period of time. Farlater, TEPCO explained the possible reason why the RCIC was operable without DC
power as follows:
The reactor water level was at the height of the nozzle for the RCIC driver line.
Two-phase water and steam flow drove the RCIC turbine like a water wheel, enablingwater injection to be maintained.
The calibrated water level coincided exactly with the height of the nozzle.
This problem was believed to be solved perfectly. Although many reappearance
analyses were published by researchers, it is presumed that most performed suchcalculations without understanding the above phenomenon.

(5)

Explosive sound in the vicinity of the suppression chamber in Unit 2
In an announcement of plant conditions at 13:00 on March 15, 2011, TEPCO
reported that explosive sound was heard near the suppression chamber in Unit 2 andthat the suppression chamber pressure simultaneously decreased. The pressure wasdescribed as "the down scale" in other reports. It was later reported that the indicator
showed 0.00 MPa [abs].
This information was later fully withdrawn. However, it led to the deductionthat a large fracture must have occurred on the suppression chamber, which caused

199

confusion in accident analyses.
This problem will be discussed in Section 7.6.
(6)

Loss of SPDSThe SPDS (Safety Parameter Display System) is a system that displays
information relating to a facility’s safety. It is an indispensable system forcomprehending plant conditions during an accident.
The SPDS is also used to assess accident conditions in ERSS, as describedabove. The SPDS lost its function with the loss of DC power. In addition, later
investigations showed that the SPDS’s data transfer component was also broken byseismic motion.
(7)
Loss of environmental monitoring
The environmental monitoring system around Fukushima Daiichi was unavailable for alength of time, which was likely caused by the loss of electrical power. Instead,environmental monitoring was performed using monitoring cars.
On the other hand, the environmental monitoring system in Fukushima Daini
remained active. In addition, air sampling was performed using an aircraft while aperiod of time. This data are very useful for the estimation of the source term and is
expected to be utilized.
As described above, many problems were experienced. Some concerned themeasurement of plant parameters; others concerned the transfer of related information.
To solve these problems, we should first recognize the differences between them. For
example:In the case that measurements cannot be obtained due to the loss of DC power:
This problem is primarily related to the electrical power system, and is not directly
related to the instrumentation system. Thus, consideration should be given to obtaining
measurements during a power loss.
In the case that water level gages require calibration: This necessity is wellunderstood by on-site operators. The gages did not fail, although the indicated values
differed from the real values.
In the case that reactor water level indicators are not reliable during severe
accidents: Although the gages did not fail, there was a problem pertaining to theavailable range of gages. This problem cannot be ignored when assessing plantconditions during a severe accident.

200

In the case that the pressure measured in the suppression chamber got different
from that of drywell in Unit 2: The problem is probably an instrumentation system
failure.
The problems concerning instrumentation systems should be categorized, and suitableapproaches should be adopted for each category.
The method for transferring information is also very important. The problems stemmingfrom instrumentation systems differed fundamentally between the categories above.
However, people besides instrumentation system experts could not easily discern thisdifference and as a result, liaisons and accident analyses were obstructed. In addition to
the transfer of raw data without delay, information should be accurately explained at thetime of information sharing.
(Drafter: Kiyoharu Abe)

201

7.6

Issues on the Updating of Information Relating to the Accident
In order to reflect the lessons learned from accidents or incidents, the first step isto understand what really happened during the accident. Inaccurate recognition of theaccident or incident may result in an impertinent reaction. Concerning the Fukushima
Daiichi accident, there is a lot of criticism on ‘what went wrong’, while only a small
number of documents describe ‘what really happened’. TEPCO has updated an
explanation of the accident situation using their investigation report. On the other hand,
many other organizations, including regulatory organizations, the Government of Japan,
and accident investigation committees do not seem to be eager in updating thedescription of the accident itself.
The Government of Japan submitted a report of the Fukushima Daiichi accident
to the IAEA Ministerial Conference on Nuclear Safety in June 2011. At that time,
however, many important facts were not found or understood. In addition, somedescriptions in the report were technically incorrect or insufficient, probably because thereport was written in confusion immediately after the accident.
This is found to be a more serious problem if we consider the internationalsociety. Other countries are also reconsidering nuclear safety in light of the Fukushima
Daiichi accident. However, even the core people of these activities sometimes
misunderstand what really happened during the Fukushima Daiichi accident. The mosttypical examples are as follows:
1) At 06:14 on March 15, an explosion was heard and then damage was found at
the roof of the fifth floor of the reactor building in Unit 4, whose fuels had been
fully extracted from the reactor. Some people estimated that the fuels in the
spent fuel pool (SFP) were exposed, and hydrogen was generated byzirconium-water reaction; therefore, the explosion occurred.
2) As introduced in Clause 7.5.3, an impulsive sound was heard near the
suppression chamber in Unit 2 around 6:00 to 6:10 on March 15. At the sametime, the pressure of the suppression chamber dropped. These reports led to theestimation that there was a large breakage in the suppression chamber.
These estimations have already been refuted. As for Example 1), hydrogen explosion
certainly occurred in the reactor building of Unit 4. However, it is concluded that thehydrogen was flowed into Unit 4 from Unit 3. It is also believed with high confidence
that the fuels of SFP in Unit 4 were never exposed throughout the accident. As for
Example 2, operators mistook the impulsive sound from Unit 4 as that occurred in the
suppression chamber of Unit 2, as already explained in Clause 7.5.3. Even today,
however, some people over the world still misunderstand that the fuel in SFP was

202

exposed in Unit 4 and that the suppression chamber was broken at the time theimpulsive sound was heard.
Why do these misunderstandings remain? During or immediately after theaccident, if some significant events such as ‘explosion’ were reported to have occurred,
the news was broadly disseminated by the press with large headlines regardless ofwhether the information was true or not. People believed the news and possibly believe
them even now. Later re-examination of accidents have been done repeatedly, andsometimes the ‘facts’ reported by the news were corrected. However, the news value of
those re-examinations is generally low. As a result, it is likely not to spread especially toforeign countries.
The two examples mentioned above are very important misunderstandings. In
addition to them, doubts, which are based on a guess, were reported widely by press,
and their investigation results are not well-known. For instance;- Pressure boundary pipes may have been breached due to seismic vibration in
Unit 1.
- Seismic capability may have been lost from the SFP in Unit 4.
In order to clarify these misunderstandings in the minds of people worldwide, weshould make efforts to repetitively explain the facts that were known in the later
examinations and that really occurred.
It is necessary to understand what really happened during the accident to learn
the lessons from the accident. In example 2), some people guessed that “there washydrogen explosion in the suppression chamber” or “this might have been caused byoverpressure because the gas phase of suppression chamber was filled with inert
nitrogen”. This misunderstanding also led to the estimation that the “decontamination
factor of suppression pool water might be not as effective as expected because theradiation level at the site boundary rose significantly just after the suppression chamber
was broken”. All of these estimations were wrong.
If we recognize the facts of accident inaccurately, the causes of the accident
would also be recognized inaccurately. As a result, measures for prevention ofrecurrence would be misdirected. In Chapters 4 and 5 of this report, we describe theevents of the accident as accurately as possible in accordance with latest findings. We
wish for all people working for nuclear safety to know the most accurate information
currently available.
Wrong theories based on such misunderstandings seem to be already sparse inJapan. However, Japan, where the accident occurred, should disseminate accurate203

information to the international society. We expect that the Government of Japan,
especially the NRA, disseminate ‘What really happened’ in the Fukushima Daiichi
accident to other countries.
(Drafter: Kiyoharu Abe)

204

7.7
Issues on the Utilization of PRA, Operating Experience, and Safety
Research
7.7.1
Relationship between PRA, Operating Experience, and Safety Research
The results of probabilistic risk assessment (PRA) or probabilistic safetyassessment (PSA), lessons learned from assessment of operating experience, and safety
research results, are intended for use in providing useful information and knowledge tomake better safety designs, safety management and regulations. Fig 7.7-1 shows thisrelationship. Operating experience and safety research results directly benefit design,
management, and regulations, and also provide foundational information for PRAs. To
give a typical example, the failure frequencies of SCCs deliver indispensable
information for a Level 1 PRA. Phenomena involved in severe accidents, themechanism of earthquake occurrence, and the distance attenuation of ground motion are
essential information for use in Level 2 PRA and seismic PRA.
Knowledge obtained from PRA, operating experience, and safety researchshould have been correctly utilized before the Fukushima Daiichi accident. After this
accident the knowledge should therefore be applied more appropriately and broadly,reflecting lessons learned from the accident.

Improvement of Safety Design and Management
Regulatory Rules, Judgment on Regulation

Deterministic Safety Assessment,
PRA

Safety ResearchInformation andKnowledge relating to
SafetyOperating Experiences

Figure 7.7-1 Utilization of PRA, Operating Experiences, and Safety Research

7.7.2

PRA
PRA systematically analyzes severe accident (SA) scenarios. It quantifies

occurrence frequencies, source terms, and health effects of SAs, and identifies

205contributing factors for the same. It also provides useful information for taking
measures to prevent SAs. Fig 7.7-2 and 7.7-3 show the methods involved in developing
a PRA, which have been widely understood by persons related to nuclear safety.
Fig.7.7-2 describes how severe accidents occur. Both internal defects in SSCs
and external impacts on SSCs can cause their failure, causing various possible effects.
Some failures cause an initiating event (IE), and some lead to the failure of the safetysystems that prevent core melt, such as the reactor shutdown system and the reactor corecooling system, or lead to the failure of safety systems that prevent containment vessel
damage, such as the containment cooling system. When both an IE and a combinationof loss of safety functions occur simultaneously, core damage and/or containment
failure can occur.
Fig.7.7-3 describes how public health is affected by a severe accident inrelation to the failure of a containment vessel. The figure illustrates release ofradioactive material into the environment (the species and amount of released
radionuclides are called the “source terms”), the diffusion of released radionuclides
under weather conditions, public exposure (which depends on emergency response
actions such as evacuation and sheltering), and finally the health effects on the public.
If we compare these scenarios illustrated in these two figures with the scenariothat actually occurred in the Fukushima Daiichi accident, we can see that the PRA
predicted quite accurately the scenario that occurred both on-site and off-site during theaccident. The PRA approach was established in the Reactor Safety Study (RSS)
published in the U.S. in 1975. Remarkable insight was shown for developing the PRA
by the pioneers.
The methodology of a stress test has also been widely applied internationallyafter the Fukushima Daiichi accident, and this is also a useful method.

Figure 7.7-2 PRA methodology (1)206Early Containment Failure ?
Late Containment Failure ?

Containment
Fracture

Releasing ofRadionuclides

Weather
Conditions

Emergency
Response Planning

Models for
Health Effects

Migration ofRadionuclides

Public Exposure

Effect for Health

Figure 7.7-3 PRA methodology (2)Therefore, PRA methodology as a whole predicted most of events that occurredin the Fukushima Daiichi accident. However, the predictability and quantitativity of
individual phenomenon were not always so high, and in relation to this during thediscussion on new safety standards of NRA, Abe gave the following comments
concerning PRA 7-64:“Considering the accuracy of PRA, for many years I have mentioned thefollowing three points as the most uncertain factors: 1) uncertainty of hazards caused byexternal events; 2) uncertainty of severe accident progression; 3) the mindset ofoperators.
In relation to the first point, I take the latest evaluation results as an example,
abstracted from the publication by the Seismic Safety Department of the JNES after theearthquake and tsunami of March 11: ’The evaluation of tsunami hazards gives ususeful information for setting the design basis of tsunami height. Taking into accountthe findings of the tsunami on March 11, the frequency of a tsunami with a wave heighthigher than 19 m is 1.5 × 10-3. This corresponds approximately to once every 670 years.
Similarly, the frequency of a tsunami higher than that of the postulated tsunami heightof about 5.7 m is 8.0 × 10-3. This corresponds approximately to once every 125 years. If
the estimation was based on previous information available for the events that took
place on 11 March, the frequency of a tsunami higher than that of the postulated heightof about 5.7 m was 3.0 × 10-3. This corresponds to approximately once every 330 years.’
The evaluations above indicate that the hazard frequency would significantly change byadding simply one more piece of tsunami data, if basic statistical processing is applied
to historical tsunami data. Here, if the frequency of occurrence of such a high tsunami
had been evaluated by a group of experts before 11 March, the result may have varied207

largely from individual to individual.
In relation to the second point, in most cases the model is prepared for eachphenomenon in a severe accident. However, it relies on the experts’ imagination tomodel what occurs when multiple phenomena or events occur simultaneously, or before
and after such events. It is also considered that some information is overlooked andomitted. For example, PRA is not able to model the event when “hydrogen leaked to thereactor building from the containment vessel and then an explosion happened there,”and “the hydrogen generated in Unit 3 leaked to the reactor building in Unit 4, which
led to the explosion in Unit 4.” Most experts thought that the suppression chamber of
Unit 2 was damaged when “a loud noise was heard in the suppression chamber of Unit
2, and then the pressure was found to be zero in the suppression chamber.” All of theseexamples indicate the insufficiency of the experts’ imagination against the progress ofsevere accident. PRA is used to roughly identify questions such as, “How often might asevere accident arise?”, “What are the key phenomena in accident sequences that result
in severe accidents?”, and “How does a severe accident progress?” However, it isdifficult to make regulations directly based on PRA result.
In addition, my personal belief is that the third point is also a major uncertainty
factor, but I declines to comment on this because of insufficient knowledge of theparticular area.”
When PRA is utilized for safety measures and regulations, its uncertainty
should always be acknowledged. In particular, in the case of a natural event, it isessential to make corresponding measures against hazards that could be in excess of
those in historic records, in addition to reflecting on the empirical data of historicaltsunamis and earthquakes in the design basis.
In Japan, although PRA has been conducted and utilized it has only been
applied to random events and earthquake. The Fukushima Daiichi accident was
triggered by an earthquake and tsunami (mainly by tsunami). In order to capture theoverall picture of risks within nuclear power plants, it is considered that an Individual
Plant Examination of External Events (IPEEE) should be executed in consideration ofwider phenomena such as tsunami, fire, and terrorist attack.
In addition to implementing PRA in a broader scope, it is also important toapply its results in the design, operation, and regulations of nuclear power plants. In
Japan (as in the U.S.), an Individual Plant Examination (IPE) was conducted tounderstand the plant-specific risk contributors, and accident management was thendeveloped on the basis of the IPE. However, there were important differences between

208the cases in the U.S. and those in Japan. In the U.S., the IPE began by analyzing internal
events (including internal floods) and then continued with an IPEEE in 1990 thatassessed external events, internal fires, high winds, external floods, seismic events, andso on. While in Japan, apart from assessing seismic events, no IPEEE was conducted
and a qualitative assessment of risk by PRA was only partly carried out.
It is also considered that researchers are partly to blame for this situation, as
they work on and refine topics that they are familiar with, but do not attempt tochallenge new fields. Due to such an omission by operators, regulators, and safety
researchers, the risk of external initiators other than seismic motion had not been
evaluated until the Fukushima Daiichi accident.
7.7.3

Operating Experience Analysis
In the light of Three Mile Island (TMI) Unit 2 nuclear accident in the U.S.,
operating experiences of abnormal events occurring in nuclear power plants are
systematically collected, analyzed, and evaluated. Lessons learned from theseoccurrences are recognized internationally and deliver important feedback forregulations and operations. Several international activities are ongoing, such as theoperating experience analysis by Institute of Nuclear Power Operations (INPO) andWorld Association of Nuclear Operators (WANO), the Incident Reporting System (IRS)
of IAEA, and the Working Group on Operating Experience (WGOE) under the umbrellaof the Committee on Nuclear Regulatory Activities (CNRA) of OECD/NEA.
In consideration of the Fukushima Daiichi accident, we need to examine
whether such activities have been effective. Has the importance of each experience beenrecognized precisely? Was the response of each event analysis appropriate? Did these
activities become empty shells?
About 10 years ago in the CNRA, the chairperson at that time (Mr. Laaksonen)insisted that the effective responses against an abnormal event should be described in
more detail than the event itself in IRS report. WGOE originally belonged to theCommittee on the Safety of Nuclear Installations (CSNI), but later, it was moved under
the CNRA. This decision was conducted with an understanding that effective responses
are more important than analysis.
With regard to the tsunami that caused the Fukushima Daiichi accident, there
were several precursory events. In 1999, parts of the Blayais nuclear power plant inFrance were flooded in relation to a combination of the tide and the rising of a river. In
2004, an earthquake occurred off the coast of Sumatra and flood events occurred in the209reactor plant of Madras, India. Industries and regulatory body in Japan analyzed thecase.
Furthermore, Japan asked IAEA an international project (Extra Budgetary
Program: EBP) collecting tsunami hazard knowledge which was not related to operating
experience analysis but accurate safety research. However, the knowledge obtained
from this project was not reflected in the operating plants.
In an international meeting, the chairperson of the IRS pointed out that “theimportance of reflecting operating experience is universally known. Analyzing andsharing information has been performed. But actually, the implementation of practices
has not been conducted well.” All participants strongly agree with his remarks. Thereappears to be a trend of ultimately only learning in relation to operating experience, and
it is essential that such experience is reflected in the actual regulations.
After the Three Mile Island accident, it was said that “this kind of accident
would never occur in Japan, because we have very high-level safety management”.
After the Chernobyl accident it was said that, “this kind of accident would never occur
in Japan, because our nuclear power plants are of a different design type.” It is evident
that such an attitude contributed to the lack of learning in Japan.
During international conferences held in relation to the Fukushima Daiichi
accident, most countries also said that “such an accident would not occur in my country
due to….” It is considered that this should not be the real response of such countries,
and that this is only a superficial official line taken. Japan, as the country responsible forthis accident, has the duty to convey accurately and repeatedly to the international
community what went wrong, and what should be improved from now on.
7.7.4

Safety Research
Nuclear safety research is intended for scientific information and to create
reasonable safety measures and regulations, with the aim of ensuring the safety of
nuclear facilities. However, among such safety research some is considered useful, butsome is considered to be simply “research for research sake,” and may not provide any
use at all.
Concerning the background of useless safety research, it can be found that
there are large discrepancies between actual safety and safety studies.
There is a saying that “a chain is no stronger than its weakest link,” andtherefore enforcing safety in relation to such “the weakest link” is considered to be
essential. In the Fukushima accident, protection against the tsunami was the weakest

210

link.
On the other hand, “something new” is absolutely indispensable forcertificating an output as a “research achievement”. The height of the achievement is
considered to be further more important than the breadth of the research, and in terms ofthis some researchers tend to try to “further strengthening the strongest link.” Other
researchers simply alter the results of their predecessors and then present such results as
something new.
There are plenty of themes involved in nuclear safety regulations, and therefore
plenty of themes to be researched. However, most cannot be solved using a single
technology, but require multiple technologies.
In this report, it has been pointed out that the nuclear safety experts were not
involved with the development of a predictive equation for tsunamis. Thus, as
previously stated, the communication between industries and the regulatory body isessential. Communication and information sharing between researchers and otherrelated parties is also important.
It is also considered that researchers should broaden their field of expertise andshould improve communication with experts in other fields to discover alternative
themes. Through this process, they will be expected to report any possible areas that
contain latent problems.
In the Seminar, the vulnerabilities of safety research, and the loss of technicalbases was also discussed. Some people said that safety research should be like a
Self-Defense Force, which means that safety research may not be useful in normal
conditions, but should be fit to serve in an accident. In other words, when severe
accidents such as that of Fukushima Daiichi or another complex event occurs,
researchers will then engage in safety research and will be able to deliver specialized
knowledge obtained from such research to support the regulatory authorities.
The staff of JNES and JAEA, the people engaged in severe accident research
after the Chernobyl and TMI accidents (and involved in PRA research) played asignificant role during the Fukushima Daiichi accident, as well as in the post-accident
analysis and the provision of necessary technical advice in response to this accident.
It is evident that experts are essential in order to advance the peaceful use of
nuclear power and its various related technologies. However, the number of experts
specializing in nuclear power has declined significantly, and this threatens the nuclear
technology infrastructure. Relevant organizations are concerned about this situation. For

211example, NISA has continued to study how to maintain the robustness of nuclear
infrastructure since its establishment. In order to encourage safety research, AESJ set up“a special technical committee of fundamental research infrastructure for light water
reactors,” to give advice on the maintenance of nuclear infrastructure technology.
Appropriate themes should be selected to conduct safety researches, and thescale of research project should be also optimized. And as a result, the existence ofpeople with high technical capabilities is maintained. Securing the human resource
acquiring high expertise is a requirement for the continual peaceful use of nuclear
power in the future.
(Draft by Kiyoharu Abe)

212

Reference
7-1
Nuclear Safety Committee. “Review Guide for Reviewing Safety Design ofLight Water Nuclear Power Reactor Facilities”. August, 1990. [in Japanese].
7-2
The Tsunami Evaluation Subcommitee, “The Nuclear Civil Engineering

7-3

7-4

7-5

7-6
7-7
7-8

7-9
7-10

Committee, Japan Society of Civil Engineers. Tsunami Assessment Method for
Nuclear Power Plants in Japan”. February, 2002. [in Japanese]. English
translation (issued in May 19, 2006) can be downloaded from JSCE’s website:
http://committees.jsce.or.jp/ceofnp/system/files/JSCE_Tsunami_060519.pdf
Nuclear Standards Committee of the Japan Electric Association. “Technicalcode for seismic design of nuclear power plants”. JEAC 4601-2008. December,2008. [in Japanese].
Nuclear Standards Committee of the Japan Electric Association. “Technicalguidelines for seismic design of nuclear power plants”. JEAG 4601-2008.
December, 2008. [in Japanese].
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Austria. January, 2012. Section 1.2~1.3. It can be downloaded from IAEA’s
website: http://www-pub.iaea.org/MTCD/publications/PDF/Pub1534_web.pdf
Guide 2, “Design consideration against natural phenomena” in reference 7-1.
[in Japanese]
Guide 27, “Design consideration against loss of power sources” in reference
7-1. [in Japanese]
Division of Nuclear Installation Safety, Department of Nuclear Safety andSecurity, International Atomic Energy Agency. “International regulatory review
service: Report to the government of Japan”. IAEA-NSNI-IRRS-2007/01.
December, 2007. The report can be downloaded at NRA’s website:
http://www.nsr.go.jp/archive/nisa/genshiryoku/files/report2.pdf
Page 39, Recommendation 9, in reference 7-8.
Abe K. private communication.

7-12

Nuclear Safety Division, Atomic Energy Society of Japan. “The minutes of 8th
Seminar to investigate the accident at the Fukushima Daiichi Nuclear Power
Station”. The University of Tokyo, Tokyo. December 19, 2012. [in Japanese].
The minutes can be downloaded from NSD’s website: http://www.aesj.or.jp/~safety/H241219seminorgiji.pdf
Page 24, in reference 7-10.

7-13

Page 24, and suggestion 1, in reference 7-8.

7-11

213

7-14
7-15
7-16

7-17

7-18

7-19
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7-21
7-22
7-23

Page 30, in reference 7-10.
Page 31, in reference 7-10.
Nuclear and Industrial Safety Subcommittee of the Advisory Committee for
Natural Resources and Energy. “Securing the safety base for nuclear energy”.
July, 2001. [in Japanese]
Central Emergency Preparedness Council. “Immediate measures for nuclear
power stations”. July, 1979. [in Japanese] This notice is reprinted at NRA’s
website: http://www.nsr.go.jp/archive/nsc/hakusyo/S56/S4-1.htm
Nuclear Safety Committee. “Emergency Preparedness and Response
Guidelines for the Vicinity of a Nuclear Power Station”. June, 1980. [inJapanse]Appendix 4, in reference 7-18.
Chapter 3, in reference 7-18.
Section 5-1, in reference 7-18.
Section 5-2, in reference 7-18.
Section 5-3, in reference 7-18.

7-24
7-25

Subsection 5-3 (1), in reference 7-18.
Organization for Economic Cooperation and Development/ Nuclear Energy
Agency, “Strategic Aspects of Nuclear and Radiologic Emergency
Management”, Paris, France. 2010. ISBN 978-92-64-99146-0. It can be
downloaded from OECD/NEA’s website: https://www.oecd-nea.org/rp/reports/2010/nea6387-strategic-aspects-of-nuclear.pdf

7-26

International Commission of Radiological Protection. “The 2007
Recommendations of the International Commission on Radiological Protection.
Publication 103”. Annals of the ICRP; 2007: 37, Nos. 2-4.
Investigation Committee on the Accident at the Fukushima Nuclear Power
Stations of Tokyo Electric Power Company. “Interim Report”. December 26,
2011. This report can be downloaded from the website of Cabinet Secretariat:
http://www.cas.go.jp/jp/seisaku/icanps/eng/interim-report.html

7-27

7-28

7-29
7-30

Investigation Committee on the Accident at the Fukushima Nuclear Power
Stations of Tokyo Electric Power Company. “Final Report”. July 23, 2012. This
report can be downloaded from the website of Cabinet Secretariat:
http://www.cas.go.jp/jp /seisaku/icanps/eng/final-report.html
Page 294-295, Subsection V 2. (1), in reference 7-27.
Page 248-250, Subsection V 2. (1), in reference 7-28.

7-31

Attachment IV-1and IV-2, in reference 7-32.

214

7-32

7-33
7-34
7-35
7-36

Nuclear Emergency Response Headquarter. “Report of Japanese Government
to the IAEA Ministerial Conference on Nuclear Safety - The Accident at
TEPCO's Fukushima Nuclear Power Stations”. June, 2011. This report can be
downloaded from the website of Prime Minister’s Cabinet:
http://japan.kantei.go.jp/kan/topics /201106/iaea_houkokusho_e.html
The monitoring results are published in “the Extension Site of Distribution
Map for Radiation Dose”: http://ramap.jaea.go.jp/map/
Page 358, in reference 7-27.
Page 354, in reference 7-27
Page 359, in reference 7-27

7-37
7-38
7-39

Page 361, in reference 7-28
Subsection 5-3 (2), in reference 7-18.
Emergency Response Technical Committee, Nuclear Safety Committee. “Theadministration of stable iodine during a nuclear emergency”. April, 2002. [inJapanese]. It can be downloaded from NRA’s website: http://www.nsr.go.jp/archive/nsc/shinsashishin/pdf/3/ho3031.pdf

7-40

The uranium processing plant’s critical accident investigation committee,
Nuclear Safety Committee. “Final Report”. December 24, 1999. [in Japanese].
It is published at NRA’s website: http://www.nsr.go.jp/archive/nsc/anzen/sonota/uran/siryo11.htm
Subsection IV. 4. (2), in reference 7-40.
Kinase S, et. al.. “Evaluation of Averted Doses to Infants by Tap Water

7-41
7-42

7-43

7-44

Restrictions after the Fukushima Daiichi Nuclear Power Plant Accident”.
Transactions of the Atomic Energy Society of Japan. 2011: 10; 149-151, [inJapanese].
Head of Water Supply Division, Health Service Beau, Ministry of Health,Labour and Welfare. “Response for water due to the Fukushima Daiichi andDaini nuclear accidents”. March 19, 2011. [in Japanese]. This publication was
also distributed at MHLW website: http://www.mhlw.go.jp/stf/houdou/2r98520000014tr1-img/2r98520000015k18.pdf
Head of Water Supply Division, Health Service Beau, Ministry of Health,Labour and Welfare. “Tap water intake by infants”. March 21, 2011. [inJapanese] This publication was also distributed at MHLW website:
http://www.mhlw.go.jp/stf/houdou/2r98520000015ox9-img/2r98520000015oyx
.pdf

215

7-45

The Ministry of Health, Labor and Welfare. “Interim report of response forwater due to radioactive contamination”. June, 2011. This report can be
downloaded from MHLW’s website: http://www.mhlw.go.jp/stf/shingi/2r9852000001g9fq-att/2r9852000001g9jp.pdf

7-46

Nuclear Emergency Response Headquarter. “Consideration of the inspection
plan, restricted items, set-up area, and release”. April 4, 2011. This publicationis also distributed at MHLW’s website : http://www.mhlw.go.jp/stf/houdou/2r98520000017tmu.html.
Subsection 5-3 (3), in reference 7-18.
International Atomic Energy Agency. “Criteria for use in preparedness and

7-47
7-48

response for a nuclear or radiological emergency : general safety guide / jointly
sponsored by the Food and Agriculture Organization of the United Nations et.

7-49al.”. IAEA-GSG-2. Vienna,

Austria. May, 2011. It can be downloaded fromthe IAEA’s website:/Pub1467_web.pdf

http://www-pub.iaea.org/MTCD/publications/PDF

Food Safety Commission. “Evaluation on radioactive materials in food”.
October, 2011. [in Japanese]. It can be downloaded from FSC’s website: http://
www.fsc.go.jp/sonota/emerg/radio_hyoka_detail.pdf

7-50

CODEX. “GENERAL STANDARD FOR CONTAMINATS AND TOXINS IN
FOOD AND FEED”. CODEX STAN 193-1995. 2009.

7-51

Working group for the revision of emergency response guides, TechnicalCommittee on Emergency Response, Nuclear Safety Committee. “Final Report”.
November, 2006. Chapter 7. [in Japanese]. This report can be downloaded from
NRA’s

website:https://www.nsr.go.jp/archive/nsc/senmon/shidai/bousin/bousin005/bousin-si005.htm

7-52

7-53

7-54

7-55

International Commission of Radiological Protection. “Application of thecommission's recommendations for the protection of people in emergency
exposure situations. ICRP Publication 109”. Annals of the ICRP 2009: 39 (1).
International Commission of Radiological Protection. “Application of thecommission's recommendations to the protection of people living in long-term
contaminated areas after a nuclear accident or a radiation emergency. ICRP
Publication 111”. Annals of the ICRP 2009: 39 (3).
International Commission of Radiological Protection. “1990 Recommendationsof the International Commission on Radiological Protection. ICRP Publication60”. Annals of the ICRP 1991: 21 (1-3).
International Atomic Energy Agency. “International basic safety standards for

216

7-56
7-57

7-58

7-59
7-60

7-61
7-62

7-63

7-64

protection against ionizing radiation and for the safety of radiation sources”.
Safety Series 115. Vienna, Austria. February, 1996.
International Commission of Radiological Protection. “Fukushima Nuclear
Power Plant Accident”. ICRP ref: 4847-5603-4313. March 21, 2011.
International Atomic Energy Agency. “Fundamental Safety Principles”. Safety
Fundamentals SF-1. Vienna, Austria. November, 2006. It can be downloaded
from IAEA’s website: http://www-pub.iaea.org/MTCD/publications/PDF/Pub1273_web.pdf
International Atomic Energy Agency. “Preparedness and Response for a
Nuclear or Radiological Emergency. Safety Requirement”. Safety Standard
Series GS-R-2. Vienna, Austria. November, 2002. It can be downloaded from
IAEA’swebsite:http://www-pub.iaea.org/MTCD/publications/PDF/Pub1133_scr.pdf
Page 10, Subsection 3.15, in reference 7-58.
Nuclear Safety Committee. “Systematization of Safety Review Guides”.
February, 2003. Page 38. [in Japanese]. It can be downloaded from NRA’s
website: http://www.nsr.go.jp/archive/nsc/shinsashishin/pdf/3/ho3004.pdf
Page 1, Subsection 1.3, in reference 7-58.
M. Saito, “Presentation files for the 5th Seminar”. The University of Tokyo,
Japan. August 20, 2012. [in Japanese]. It can be downloaded from NSD’s
website: http://www.aesj.or.jp/~safety/H240820seminorsiryou4.pdf
Nuclear Safety Division, Atomic Energy Society of Japan. “The minutes of 5th
Seminar to investigate the accident at the Fukushima Daiichi Nuclear Power
Station”. The University of Tokyo, Tokyo. August 20, 2012. [in Japanese]. The
minutes can be downloaded from NSD’s website: http://www.aesj.or.jp/~safety/H240820seminorgiji.pdf
Abe K. “Comments from experts to the document of 5th meeting on discussing
new regulatory standards for light water reactor power plants”. Nuclear
Regulatory Authority. December 13, 2012. [in Japanese]. It can be downloaded
from
NRA’s
website:https://www.nsr.go.jp/committee/yuushikisya/shin_anzenkijyun/data/0006_08.pdf

217

8.

Conclusions
The NSD held eight open seminars concerning the Fukushima Daiichi accident
to share information and to improve discussion. Each member of the NSD is an expert
in nuclear safety. As most of us are closely involved with the accident, we gathered inan academic society to deeply discuss issues on a technical basis. The specific points fordiscussion were set in the “Seminars to Investigate the Accident at the
Fukushima-Daiichi Nuclear Power Station,” (the Seminar). Common problems and thebackground issues were categorized from the individual events which occurred in theaccident. Challenges for securing safety were then extracted based on reflections on theaccident.
The objective of securing safety in nuclear facilities is to protect people and theenvironment from harmful effects of ionizing radiation. To achieve this objective, adesign basis with wide margins should be established. Since a nuclear power plant is anextremely complex artificial system, we need to make a conscious effort to take apanoramic view of whole system. Complex systems are structured by various
technologies which belong to the different specific areas. A gap between such technicalfields leads to a defect of the system and may become the initiator of an accident.
Certain principles and approaches for securing nuclear safety have beenutilized for many years. Unvarying principles should be elucidated, in addition toadvancing new items and revisions in light of the Fukushima Daiichi accident.
(1) DiD to secure safety
DiD is an unvarying principle that is to be used after the Fukushima Daiichi
accident. The risk of a severe accident beyond the design basis has been recognized, andAM measures have been prepared. The Fukushima Daiichi accident showed that theprotection against natural events, especially tsunamis, was insufficient. Natural events
beyond the design basis can cause simultaneous failures of multiple SSCs, and hugehazards are possible, even if they have extremely low frequencies.
During the Fukushima Daiichi accident, many SSCs with a safety function losttheir function at the same time because of inundation by a tsunami wave exceeding thedesign hypothesis. As the result, the reactor core became molten and a huge amount ofradioactive materials was released to the environment.
The risk of severe accident initiated by natural events is usually higher than the
one initiated by random failures. In the case of seismic motion, this fact was indicatedby the seismic PRA, and therefore the Seismic Design Guide was improved before the
Fukushima Daiichi accident. However, independence between layers of DiD related to218natural events was not prepared sufficiently. Many individual problems were related tothis, for example an insufficient design basis against external initiators, the diversity of
injection systems and cooling systems, the inefficiency of AM measures, and theenforcement of emergency protection operations. It is a matter of great regret to us that
we omitted such important principles.
Other nuclear power stations avoided severe accident by the use of various
measures, such as an initial decision to elevate the site with the intention of avoiding therisk of tsunami, and a gradual review of postulated tsunami height based on renewed
knowledge.
Another important point is the lack of communication between specialized
areas concerning the risk of postulated tsunami for a design basis. The objective for
deciding a hypothesis should be to secure nuclear safety, and this basic goal was notshared by the relevant people. Nuclear safety experts should have initiated intensive
communication with tsunami experts. The design basis hazard needed to be proposed
quantitatively, and the corresponding tsunami height should have been decided based on
deep communication. A review of standards needs to be based on such discussions.
(2) Severe Accident Measures and Securing their Efficiency
There is some difficulty in setting protection levels against natural hazards andimplementing safety measures corresponding to the levels. It is hard to cover all
possible scenarios to propagate beyond the design basis. Flexible and elastic responses
are effective, such as the utilization of portable equipment. Moreover, organizations andindividuals should be highly able to make decisions and to take leadership.
The difficulty of operating AM measures under a severe accident condition was
also recognized. Restrictions of time and space existed when the operation was
performed under a high radioactive environment. The control room was uninhabitable.
Some AM measures could not be performed easily under an actual severe accident
condition. We would like to point out that prevention measures against severe accidents
are extremely important even though the mitigation measures should of course be
improved.
DiD should be used as an effective principle for nuclear safety, and thisrequires the collaboration of many factors. For instance, safety should be secured bymultiple and multilayered measures both on-site and off-site. In addition, measures tosecure the effectiveness of AMs should be considered, and should be reviewed
sustainably.

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(3) Nuclear Emergency Response
There was confusion in relation to the emergency response after the Fukushima
Daiichi accident. However, evacuation and sheltering were effectively operated basedon the site condition and were gradually extended to 3, 10, and 20 km, without
dependence on prediction systems such as ERSS and SPEEDI.
Being sufficiently prepared for an emergency is necessary. The emergency
response needs to be operated based on IAEA’s protection plan, which was establishedin accordance with the lessons from the accidents at TMI-2 and Chernobyl.
Standardization based on the technical assessments of practices needs to be utilized.
The roles of each participant need to be clarified along with emergency
management. The chain of command and the assignment of duty should be gradually
established. A bi-directional scheme of decision making and crisis communication isrequired. Decisions should be made based on the judgment of experts and theinformation gathered. Emergency protective actions including distribution of
instructions and warnings should be managed systematically along the scheme.
(4) Safety Research for the Future
To secure safety by DiD, all possible initiators should be examined. Risk
assessment needs to show the whole picture, which would therefore improve therequirements for SSCs design, management of SSCs, and appropriate AM.
It is possible to understand the propagation of each individual event during the
Fukushima Daiichi accident using PRA. PRA had been applied only to random eventsand seismic motion. Since a tsunami was described as an event associated with anearthquake, structuring of specific scenarios for assessing tsunamis was not improved,
although the effects of tsunami event propagation are considerably different from thoseof seismic hazards. The research should be extended to challenges regarding events
caused by other initiators, for example tsunami, fire, and terrorist attack.
A researcher tends to be engaged in one subject as deep as possible, and his/her
effort is usually concentrated on his/her most favorite technical area. In contrast, safety
disappears through a gap existing between expertise and research fields. When a newresearch plan is considered, the challenges involved should use a panoramic view. The
use of safety research is indispensable, and achievable objects of nuclear safety should
be discussed. A map of technical challenges should be prepared by squarely facing thereal capacity of current technology. In addition, load maps to achieve these challenges
should be proposed with consideration of both the short term and the long term. The
above approach is essential in safety research and for continuous improvement.

220

Safety research is expected to indicate latent problems and to warn of such,based on renewed scientific and/or technological knowledge. All industries, academia,
and the government have a duty to promote safety research through frank discussions
and information sharing.
(5) System of Nuclear Safety Regulation
Scientific rationality is a key to achieving nuclear safety and trust from thepublic. A graded approach, which assigns resources based on risk, is required both by
the regulatory body and industries. Risk informed regulation should be adopted. A
global assessment of latent defects related to important measures and the oversight ofsignificant risk is more important than making efforts to examine a topical risk in detail.
To achieve this objective, performance regulations should be adopted. The
regulatory body should decide the requirements for performance without help fromother bodies. Then, the codes and standards established by academic societies should be
adopted as the specification codes for regulation after a strict review. Collaboration of
all industries, academia, and government should then be achieved simultaneously withsecuring independency of regulation.
It is considered necessary to point out again that human resources are importanton a national basis. The government should maintain its crisis management ability, andshould also collect and evaluate a knowledge basis required to make a global decision
on regulation. The basis of human resources should include development systems,
evaluation systems, and a personnel management system for experts.
(6) Communication
Communication is commonly related to the above problems. Nuclear power iscategorized as an integrated engineering, and its research target is huge and complex
nuclear plant systems. Engineering judgments based on communication between people
with different technical backgrounds are required to achieve a risk analysis, which isneeded to perform a probabilistic safety assessment. As clarified by the lack of essential
communication and discussions when the postulated tsunami height was decided,common understanding should be shared by experts to achieve nuclear safety. Experts
directly related to nuclear safety should play a key role to lead the communication.
Bi-directional communication between experts, utilities, and the regulatory
body is necessary to achieve nuclear safety objectives. In addition, information transfer
and bi-directional communication sharing including all participants relating to an
emergency should be examined, and all involved should be given advanced training.

221

Such participants should include governmental ministries, members of the police, fire
defense, the Self Defense Force, local government, residents, and the press.
(7) Responsibility of the Nuclear Safety Division
It is the responsibility of the NSD to indicate the basic concepts for securing
nuclear safety and to share such concepts. The DiD approach is still important after the
Fukushima Daiichi accident and should be improved continuously by both utilities andregulatory bodies. Regulation using a scientific rationality enhances the improvement.
It is important for regulatory body to keep its independency, which will be secured byappropriate

collaboration

betweenindustries,

academia,

andgovernment.

Organizations and activities for securing nuclear safety should never lose substance.

Regulations should never become a mere façade. It often occurs that the preparation ofdetail manuals for promoting safety culture leads the loss of the validity of activities.
Academic society gathers opinions and the latest knowledge. Individuals
belonging to both the regulatory body and utilities are joining the NSD. A relationshipof mutual trust should be developed by communication between related people.
The NRA issued the framework of a new safety regulation in February, 2013.
All relevant people are required to behave in a responsible manner with a proactive
mind. International discussion and standards should be absorbed into all individual’s
activities. The important challenges described in this report should also be publishedand reported to international society to enhance discussion.
The NSD continues to talk about the standard of conduct to counter risks. We
are continuing to find the essential challenges involved in achieving nuclear safety, andare improving the effectiveness of activities concerning nuclear safety. We are also
continuing to play the role of enhancing a cross-sectional approach between wide
technical areas involved in nuclear power.
(Drafter: Naoto Sekimura)

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9.

Final Statements
More than two years have passed since the Fukushima Daiichi accident causedby the Tohoku-Pacific Ocean Earthquake on 11 March 2011. This accident was analysedby domestic and foreign organizations. Several lessons were learned, and proposals
were made. Some of them have now been put into practice. The NRA has been
established as an independent organization, and is preparing the framework of Nuclear
Emergency Guides and regulatory requirements for severe accident management.
Nuclear power operators are incorporating emergency safety measures. The measuresand their effectiveness are reviewed by stress tests. Other improvement measures
proposed by NISA, the previous regulatory body, are also being incorporated.
As the gathering of the experts who bear responsibility for the accident, theNSD conducted a series of seminars—‘Seminars to investigate the accident at the
Fukushima Daiichi Nuclear Power Station’ since February 2012. Eight seminars wereorganized through almost a year. The experts who have been working on plant design,operation, safety regulation, and safety research reviewed the accident and proposed thechallenges based on their reflections. The content covered many topics, for instance, theinsufficient assumption of external initiators, insufficient design of DiD, lack of fluent
communication, and insufficient preparedness for a nuclear emergency. The participants
were extremely interested in these topics, and opinions and questions were exchanged
between lecturers and participants. Some of the important questions discussed were as
follows: What preparations were made for taking measures against severe accidents?
What assumptions did the plant designers and owners make from the viewpoints ofsevere accident? How did they prepare for such an accident? Which measures wereinsufficient? What did the operators on the site think, and how did they act? What were
the difficulties they faced? How did engineers other than nuclear power engineers feel
about nuclear power? Which regulatory body focused on this incident?
We hope that people related to nuclear energy read the records of our
discussion and utilize the results in their respective organizations. We also hope that thisreport will contribute to a better understanding of accident progression by general public.
We expect this report to be used in conjunction with other reports to improve the levelof nuclear safety through the efforts of the parties concerned.
We would like to appreciate all participants of the seminar series, especiallythose who managed the preparation, delivered lectures, and participated in the serious
discussions.
(Drafter: Takashi Nitta)

223

Appendix 1: List of Abbreviations
AESJ

The Atomic Energy Society of Japan

AM

Accident Management

AMG

Accident Management Guideline

CAMS

Containment Atmospheric Monitoring System

CS

Core Spray System

CST

Condensate Storage Tank

DD-FP

Diesel Driven Fire Pump

DiD

Defense in Depth

DW

Dry Well

EDG

Emergency Diesel Generator

EECW

Emergency Equipment Cooling Water System

EOP

Emergency Operation Procedure

FP

Fire Protection System

HERP

The Headquarters for Earthquake Research Promotion

HPCI

High Pressure Coolant Injection System

HPCS

High Pressure Core Spray System

IC

Isolation Condenser

INES

International Nuclear and Radiological Event ScaleJSCE

The Japan Society of Civil Engineers

LOCA

Loss of Coolant Accident

LPCI

Low Pressure Core Injection System

M/C

Metal-Clad Switch Gear

MCCI

Molten Core Concrete Interaction

MSIV

Main Steam Isolation Valve

MUWC

Make-Up Water System

NRA

Nuclear Regulatory Authority

NSC

Nuclear Safety Commission

P/C

Power Center

PCV

Primary Containment Vessel

PRA

Probabilistic Risk Assessment

RCIC

Reactor Core Isolation Cooling System
224

RHR

Residual Heat Removal System

RHRC

RHR Cooling Water System

RHRS

RHR Sea Water System

RPV

Reactor Pressure Vessel

S/C

Suppression Chamber

S/P

Suppression Pool

SBO

Station Blackout

SFD

Self-Defense Force

SFP

Spent Fuel Pool

SGTS

Stand-by Gas Treatment System

SRV

Safety Relief Valve

TAF

Top of Active FuelTEPCOTokyo Electric Power Company

USNRC

The United States Nuclear Regulatory Commission

WW

Wet Well225

Appendix 2: Contributors’ List
Chair person

Kiyoharu Abe

Japan Nuclear Energy Safety OrganizationVice-chair

Takashi Nitta

The Japan Atomic Power Company

persons

Naoto Sekimura

Excutive
Committee
Members

Participants+

The University of Tokyo

Toyoshi Fuketa
Kohei Hisamochi
Toshimitsu Homma
Masanori Ino
Chikako Iwaki

Japan Atomic Energy Agency
Hitachi-GE Nuclear Energy, Ltd.
Japan Atomic Energy Agency
Toshiba Corporation
Toshiba Corporation

Nubiry Kubo
Tsutomu Matsui

Japan Atomic Energy Agency
Chubu Electric Power Co., Inc.

Shingo Matsuoka
Koichi Miyata
Ken Muramatsu
Takehiko Nakamura*

Japan Nuclear Fuel Limited.
Tokyo Electric Power Company
Tokyo City University
Japan Atomic Energy Agency

Narumitsu Umezawa
Shigeru Urata

Mitsubishi Heavy Industries, LTD.
The Kansai Electric Power Co., Inc.

Hiroshi Miyano
Kumiaki Moriya
Takanori Tanaka

Hosei University
Hitachi-GE Nuclear Energy, Ltd.
The Institute of Applied Energy

Contributors for Yasumitsu Fujii

The Kansai Electric Power Co., Inc.

Drafting

Tohoku-Electric Power Co., Inc.
Tokyo Electric Power Company
Tokyo Electric Power Company
Tohoku-Electric Power Co., Inc.
The Japan Atomic Power Company

Kosuke Kato
Atsushi Kawamoto
Kazuyuki Nagasawa
Tatsuyuki Sakuraba
Noriaki Tomura

Contributors for Kenta Murakami
Translation
Liu LiThe University of Tokyo
The University of Tokyo*

226

from September, + until September

List of Draft Writers
1. Introduction

Kiyoharu Abe

2. Outline of the Seminar Series and Glossary of TechnicalTerms
2.1 Outline of the Seminar Series
2.2 Wording

Kiyoharu Abe
and Ken Murakamtsu
Kiyoharu Abe

3. Summary of the Safety Design of Fukushima Daiichi
3.1 Safety Design of the Facility

Kumiaki Moriya

3.2 Seismic Design and Tsunami-Proof Design

Kumiaki Moriya

4.

Tsunami Triggered by the Tohoku-Pacific Ocean Noriaki Tomura,
Earthquake and Histories to Determine Postulated Kazuyuki Nagasawa,
Tsunamiand Koichi Miyata

5. Summary of Fukushima Daiichi Nuclear Accident
5.1 Damage by Earthquake and Tsunami

Koichi Miyata

5.2 Fukushima Daiichi Unit 1
5.3 Fukushima Daiichi Unit 2
5.4 Fukushima Daiichi Unit 3
5.5 Spent Fuel Pool
5.6 Fukushima Daiichi Unit 5, 6
5.7 Evaluation of the Released Radioactive Material

Koichi Miyata
Koichi Miyata
Koichi Miyata
Koichi Miyata
Koichi Miyata
Atsushi Kawamoto

5.8

Physical Distribution of Resources during the Toshimitsu Homma
Accident
5.9 Off-site Emergency Response
Kiyoharu Abe
5.10 INES
Kiyoharu Abe,and Naoto Sekimura
6. Summary of Other Nuclear Power Plants
6.1 Fukushima Daini Nuclear Power Station
6.2 Onagawa Nuclear Power Station
6.3 Tokai No. 2 Nuclear Power Station
6.4 Comparison of Events in Nuclear Power Plants

227

Kazuyuki Nagasawa
Tatsuyuki Sakuraba,
and Kosuke Kato
Noriaki Tomura
Yasumitsu Fujii,
and Kiyoharu Abe

7. Issues Identified by the Accident
7.1 Summary of Issues According to DiD Concept
7.2 Issues on Safety Design
7.3 Issues on Safety Regulation
7.4 Issues on Nuclear Emergency Response
7.5 Issues on Cooperation and Communication during the
Accident
7.6 Issues on the Updating of Information relating to theAccident
7.7

Kiyoharu Abe
and Kumiaki Moriya
Kumiaki Moriya
Kiyoharu Abe
Toshimitsu Homma
Kiyoharu Abe
Kiyoharu Abe

Issues on the Utilization of PRA, Operating Kiyoharu Abe
Experience, and Safety Research

8. Conclusions

Naoto Sekimura

9.

Takashi Nitta

Final Statement

EditorTsutomu Matsui

English Translation

Kenta Murakami
Liu Li

228