Fukushima Daiichi Accident – Technical Causal Factor … Technical...10CFR50 Appendix B, 10CFR...

2012 TECHNICAL REPORT

Electric Power Research Institute 3420 Hillview Avenue, Palo Alto, California 94304-1338 • PO Box 10412, Palo Alto, California 94303-0813 USA

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Fukushima Daiichi Accident – Technical Causal Factor Analysis

EPRI Project Manager K. Huffman

ELECTRIC POWER RESEARCH INSTITUTE

3420 Hillview Avenue, Palo Alto, California 94304-1338 ▪ PO Box 10412, Palo Alto, California 94303-0813 ▪ USA 800.313.3774 ▪ 650.855.2121 ▪ [email protected] ▪ www.epri.com

This document does NOT meet the requirements of 10CFR50 Appendix B, 10CFR Part 21,

ANSI N45.2-1977 and/or the intent of ISO-9001 (1994)

Fukushima Daiichi Accident –Technical Causal Factor Analysis 1024946

Final Report, March 2012

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Fukushima Daiichi Accident – Technical Causal Factor Analysis. EPRI, Palo Alto, CA: 2012. 1024946.

iii

ACKNOWLEDGMENTS

The following organizations prepared this report:

Electric Power Research Institute (EPRI) 1300 West W.T. Harris Blvd. Charlotte, NC 28262

Principal Investigator K. Huffman

Lucius Pitkin Inc. 304 Hudson St. New York, NY 10013

Principal Investigators P. Bruck T. Esselman P. Streeter

This report describes research sponsored by EPRI.

v

ABSTRACT

On March 11, 2011, the Fukushima Daiichi nuclear power plant experienced a seismic event and subsequent tsunami. The accident and the ensuing mitigation and recovery activities occurred over several days, involved a number of incidents, and might provide several opportunities for lessons learned. The objective of this report is to determine the fundamental causative factors for the loss of critical systems at the Fukushima Daiichi reactors that resulted in core damage and subsequent radioactive release. The analytical approach used structured event information to support the identification of causal factors; this information can be used in root cause analysis and corrective action determinations.

The technical analysis traced the cause for the eventual loss of all practical cooling paths for the reactors to the tsunami’s flooding of the plant protection. Specifically, the analysis identified the significant difference between the design basis tsunami height and the actual tsunami height, as well as the limitations of beyond-design-basis tsunami protection or mitigation that could address the effects of the actual event. From a causal analysis perspective, these were caused by a methodology that specified that the rupture of combinations of geological fault segments in the vicinity of the plant need not be considered in establishing the design basis tsunami height. The tsunami that occurred was caused by a combined rupture of multiple fault segments.

Keywords Earthquake Fukushima Daiichi Tsunami

vii

CONTENTS

1 INTRODUCTION .................................................................................................................... 1-1

Background ........................................................................................................................... 1-1

2 ACCIDENT TIMELINES ......................................................................................................... 2-3

Unit 1 ..................................................................................................................................... 2-4

Unit 2 ..................................................................................................................................... 2-6

Unit 3 ..................................................................................................................................... 2-9

3 SEISMIC AND TSUNAMI DESIGN BASES ........................................................................... 3-1

Seismic .................................................................................................................................. 3-1

Tsunami ................................................................................................................................ 3-4

4 MARCH 11 SEISMIC EVENT ................................................................................................. 4-1

5 MARCH 11 TSUNAMI ............................................................................................................ 5-1

6 PLANT CAPABILITIES AND RESPONSE ............................................................................ 6-1

Operational Capabilities and Response ................................................................................ 6-1

Seismic Capabilities and Response ...................................................................................... 6-2

Tsunami Capabilities and Response ..................................................................................... 6-2

7 CAUSAL ANALYSIS .............................................................................................................. 7-1

Plant Protection ..................................................................................................................... 7-1

Actual Event .......................................................................................................................... 7-1

Comparison of Design Basis and Actual Event ..................................................................... 7-1

Causal Analysis ..................................................................................................................... 7-2

8 CONCLUSIONS ..................................................................................................................... 8-1

9 REFERENCES ....................................................................................................................... 9-1

ix

LIST OF FIGURES

Figure 1-1 Arrangement of Fukushima Daiichi Units ................................................................. 1-2

Figure 2-1 Flow Diagram of Cooling Systems, Unit 1, with ECCS Highlighted ........................ 2-14

Figure 2-2 Flow Diagram of Cooling Systems, Units 2 and 3, with ECCS Highlighted ............ 2-15

Figure 2-3 High Pressure Coolant Injection (HPCI) Flow Diagram, Units 1 – 3 ....................... 2-16

Figure 2-4 Isolation Condenser (IC) Flow Diagram, Unit 1 ...................................................... 2-17

Figure 2-5 Reactor Core Isolation Cooling (RCIC) Flow Diagram, Units 2 and 3 .................... 2-18

Figure 2-6 Residual Heat Removal Flow Diagram, Units 2 and 3 ........................................... 2-19

Figure 2-7 Reactor Shutdown Cooling (SHC) Flow Diagram, Unit 1 ....................................... 2-20

Figure 2-8 Main Steam Safety Relief Valve (MSRV) Flow Diagram, Unit 1 ............................. 2-21

Figure 2-9 Main Steam Safety Relief Valve (MSRV) Flow Diagram, Units 2 and 3 ................. 2-22

Figure 2-10 DC Power Supply System, Unit 1 ......................................................................... 2-23

Figure 2-11 DC Power Supply System, Units 2 and 3 ............................................................. 2-24

Figure 3-1 Comparison of Original Seismic Spectra with Revised Seismic Spectra .................. 3-3

Figure 4-1 Measured Horizontal East-West Seismic Spectra Compared to the Design Basis Spectra for Units 2, 3, and 5 ..................................................................................... 4-2

Figure 5-1 Plan View of Site (Units 1 to 4 Only) Showing Flooded Regions Following the Tsunami ............................................................................................................................. 5-2

Figure 5-2 Section View (Approximate) through Unit 1 Showing Flood Depth .......................... 5-3

Figure 5-3 Photos at Points A and B in Figure 5-1 .................................................................... 5-3

Figure 5-4 Photos at Point C Showing Flooding Progression .................................................... 5-4

Figure 5-5 Photos Showing Seawall During the Tsunami .......................................................... 5-5

Figure 7-1 Causal Chart ............................................................................................................. 7-3

xi

LIST OF TABLES

Table 1-1 Comparison of Units 1 to 6 ........................................................................................ 1-1

Table 2-1 Units 1 - 3 Power Source Unavailability Following the Earthquake/Tsunami .......... 2-12

Table 2-2 Status of Electrical Supply after Tsunami ................................................................ 2-13

Table 4-1 Maximum Measured Acceleration Values Compared to the Design Basis Acceleration Values ........................................................................................................... 4-1

Table 6-1 Timeline of Key Cooling Systems Failures ................................................................ 6-1

1-1

1 INTRODUCTION

On March 11, 2011, the Fukushima Daiichi nuclear power plant experienced a seismic event and subsequent tsunami. This event has been broadly reported [1 - 6].

The objective of this report is to determine the fundamental causative factors for the loss of critical systems at the Fukushima Daiichi reactors that resulted in core damage and subsequent uncontrolled release of radioactive materials to the environment. An events and causal factors analysis approach was used that structures event information and supports identification of causal factors that can then be used in root cause and corrective action determinations [13]. The purpose for determining these causal factors is to ensure that: (1) the most important factors that caused the accident are highly evident; and (2) this knowledge will support identification of safety improvements that are meaningful and appropriate. This report accomplishes the first of these two purposes.

The plant response after the loss of critical systems at Fukushima Daiichi is not described here because the intent of this report is to identify the factors that led to loss of critical systems that then resulted in core damage. Because only Units 1 through 3 were operating at the time of the events, the primary focus of this report is on Units 1 through 3.

Background Fukushima Daiichi consists is located along the northeast coast of Japan, bordering the western edge of the Pacific Ocean. Initial siting occurred between 1967 and 1973, with each of the six units coming on-line between 1971 and 1979. Operational startup dates, power output, and general design information are shown in Table 1-1 [2].

Table 1-1 Comparison of Units 1 to 6

Unit Startup MWe Output Reactor Type, Containment High Pressure Cooling*

1 1971 460 BWR-3, Mark I IC, HPCI

2 1974 784 BWR-4, Mark I RCIC, HPCI




6 1979 1100 BWR-5, Mark II RCIC, HPCS

* IC: Isolation Condenser HPCI: High Pressure Coolant Injection RCIC: Reactor Core Isolation Cooling HPCS: High Pressure Core Spray

Introduction

1-2

The plant is designed such that grouped units share some common facilities and structures, such as control rooms, and vent stacks. This commonality applies to Units 1 and 2, Units 3 and 4, and Units 5 and 6.

In addition to individual unit spent fuel pools, the plant also has a shared spent fuel pool and a shared dry cask storage facility. The shared pool and the dry cask storage are for all six units, with older spent fuel having been transferred from the individual unit pools to the shared pool or dry cask storage on a periodic basis. The shared spent fuel pool is located on the inland side (west) of Unit 4. The dry cask storage facility is located between Units 1 and 5 along the coast.

The general relation between the units prior to the earthquake and the tsunami is shown in Figure 1-1.

Figure 1-1 Arrangement of Fukushima Daiichi Units

2-3

2 ACCIDENT TIMELINES

An overview of the events as they occurred is presented here, noting only those items considered of significance relative to the times when critical systems were lost. The overview of the events provides the timeline on a per Unit basis. Within each unit, the timeline is further broken down by system, as appropriate, with key events/actions indicated. The timelines are not inclusive, but focus on systems that could provide cooling functions. The timelines continue until the safety systems become unavailable. Flow diagrams extracted from [1] are included as Figures 2-1 to 2-9 and these apply to Units 1, 2, and 3 as identified on each figure. Systems identified as being with the emergency core cooling system (ECCS) are identified on the overall cooling system flow diagrams in Figures 2-1 and 2-2.

The time of the first seismic ground motion is considered the baseline and all subsequent events are identified both in real plant time and differential time from this baseline. Timeline data are taken from [1], [4], [5], and [6]. Equipment elevation data (where identified) are taken from [1]. For reference, all elevations are given with respect to the base tide level used for original construction and are identified as “O.P.” referring to Onahama Peil (level). Thus, the base water level is O.P. 0.0m and site grade for Units 1 to 3 is approximately O.P. +10.0m.

A key to understanding the inability of critical systems to function is the unavailability of power sources following the earthquake and tsunami. To this end, information from [1], [5] and [6] are summarized in Table 2-1. This table describes the unavailability causes of the normal site power sources for each unit. Table 2-2 extracted from [6] repeats some of the information from Table 2-1 but in a different fashion. Figures 2-10 and 2-11 show DC power supply diagrams for Unit 1 and Units 2/3, respectively extracted from [1].

All off-site AC power was lost as a result of the earthquake. The loss of on-site AC power was caused by the submergence of the emergency diesel generators and electrical distribution system equipment inside the plants. Loss of DC power in Units 1 and 2 was caused by submergence of electrical distribution system equipment. The loss of DC power in Unit 3 to some systems was caused by submergence of electrical distribution system equipment and then eventually by full discharge of the batteries.

Accident Timelines

2-4

Unit 1 UNIT 1: ISOLATION CONDENSER (IC) (location: Reactor Building (RB) 4th floor, elev = O.P. +31.00m) Date Time Diff (hrs) Event/Action 3/11/2011 1446 0.00 seismic ground motions exceed setpoints 3/11/2011 1446 0.00 scram signal initiated, control rods inserted 3/11/2011 1447 0.02 Emergency diesel generators (EDGs) start on loss of

offsite power 3/11/2011 1452 0.10 IC's start automatically 3/11/2011 1503 0.28 IC's manually removed from service 3/11/2011 1510 0.40 IC(A) restarted 3/11/2011 1519 0.55 IC(A) stopped 3/11/2011 1524 0.63 IC(A) restarted 3/11/2011 1526 0.67 IC(A) stopped 3/11/2011 1527 0.68 first tsunami hits 3/11/2011 1532 0.77 IC(A) restarted 3/11/2011 1534 0.80 IC(A) stopped 3/11/2011 1535 0.82 second tsunami hits 3/11/2011 1537 0.85 loss of all AC power 3/11/2011 1537 0.85 loss of all DC power 3/11/2011 1537 0.85 valve status of IC lost on control panel

3/11/2011 1818 3.53 partial DC power returned, IC(A) placed in service (steam observed from condenser)

3/11/2011 1825 3.65 IC(A) removed from service 3/11/2011 2130 6.73 IC(A) placed in service (steam observed)

3/12/2011 0148 11.03 IC(A) stopped (diesel-driven fire protection pump failed)

UNIT 1: HIGH PRESSURE CORE INJECTION (HPCI) (location: RB basement, elev = O.P. -1.23m) Date Time Diff (hrs) Event/Action 3/11/2011 1446 0.00 seismic ground motions exceed setpoints 3/11/2011 1446 0.00 scram signal initiated, control rods inserted 3/11/2011 1447 0.02 EDGs start on loss of offsite power 3/11/2011 1527 0.68 first tsunami hits 3/11/2011 1535 0.82 second tsunami hits 3/11/2011 1537 0.85 loss of all AC power 3/11/2011 1537 0.85 loss of all DC power 3/11/2011 1537 0.85 HPCI determined to be non-functional

Accident Timelines

2-5

UNIT 1: OTHER HIGH PRESSURE SYSTEMS -- STANDBY LIQUID CONTROL (SLC), CONTROL ROD DRIVE (CRD) WATER (location SLC: unknown) (location CRD: unknown) Date Time Diff (hrs) Event/Action 3/11/2011 1446 0.00 seismic ground motions exceed setpoints 3/11/2011 1446 0.00 scram signal initiated, control rods inserted 3/11/2011 1447 0.02 EDGs start on loss of offsite power 3/11/2011 1527 0.68 first tsunami hits 3/11/2011 1535 0.82 second tsunami hits 3/11/2011 1537 0.85 loss of all AC power 3/11/2011 1537 0.85 loss of all DC power 3/11/2011 1537 0.85 SLC, CRD non-functional due to loss of AC power

UNIT 1: OTHER POTENTIAL COOLING SYSTEMS – CONTAINMENT COOLING (CCS), CLOSED COOLING SEA WATER (CCSW), MAKEUP WATER (MUWC), FIRE PROTECTION, SPENT FUEL POOL (SFP) COOLING (location CCS: RB basement, elev = O.P. -1.23m) (location CCSW: outdoors, elev = O.P. +4.00m) (location MUWC: TB basement, elev = O.P. +3.20m) (location fire protection: unknown) (location SFP: RB 3rd floor, elev = O.P. +25.90m) Date Time Diff (hrs) Event/Action 3/11/2011 1446 0.00 seismic ground motions exceed setpoints 3/11/2011 1446 0.00 scram signal initiated, control rods inserted 3/11/2011 1447 0.02 EDGs start on loss of offsite power 3/11/2011 1527 0.68 first tsunami hits 3/11/2011 1535 0.82 second tsunami hits 3/11/2011 1537 0.85 loss of all AC power 3/11/2011 1537 0.85 loss of all DC power

3/11/2011 1537 0.85 CCS, CCSW, MUWC, SFP, motor-driven fire protection pump non-functional due to loss of AC power

Accident Timelines

2-6

Unit 2 UNIT 2: REACTOR CORE ISOLATION COOLING (RCIC) (location RCIC: RB basement, elev = O.P. -2.06m) Date Time Diff (hrs) Event/Action 3/11/2011 1446 0.00 seismic ground motions exceed setpoints 3/11/2011 1447 0.02 scram signal initiated, control rods inserted 3/11/2011 1447 0.02 EDGs start on loss of offsite power 3/11/2011 1450 0.07 RCIC started by operators 3/11/2011 1451 0.08 RCIC tripped due to high reactor water level 3/11/2011 1502 0.27 RCIC started by operators 3/11/2011 1527 0.68 first tsunami hits 3/11/2011 1528 0.70 RCIC tripped due to high reactor water level 3/11/2011 1535 0.82 second tsunami hits 3/11/2011 1539 0.88 RCIC started by operators 3/11/2011 1541 0.92 loss of all AC power 3/11/2011 1541 0.92 loss of all DC power 3/11/2011 1541 0.92 RCIC operational status unknown

3/11/2011 2102 6.27 RCIC operational status still unknown; no reactor water level indication

3/12/2011 0030 9.73 RCIC apparently still running based on walkdown (sounds only)

3/12/2011 0255 12.15 RCIC apparently still running based on RCIC discharge pressure being high

3/12/2011 0500 14.23 RCIC water source changed from condensate storage tank to suppression chamber

3/12/2011 2100 30.23 RCIC still operating

3/13/2011 1040 43.90 RCIC still operating 3/13/2011 1350 47.07 RCIC still operating

3/14/2011 1325 70.65 RCIC assumed failed based on lowered reactor water level

Accident Timelines

2-7

UNIT 2: HIGH PRESSURE CORE INJECTION (HPCI) (location HPCI: RB basement, elev = O.P. -2.06m) Date Time Diff (hrs) Event/Action 3/11/2011 1446 0.00 seismic ground motions exceed setpoints 3/11/2011 1447 0.02 scram signal initiated, control rods inserted 3/11/2011 1447 0.02 EDGs start on loss of offsite power 3/11/2011 1527 0.68 first tsunami hits 3/11/2011 1535 0.82 second tsunami hits 3/11/2011 1541 0.92 loss of all AC power 3/11/2011 1541 0.92 loss of all DC power 3/11/2011 1541 0.92 HPCI determined to be non-functional

UNIT 2: OTHER HIGH PRESSURE SYSTEMS -- STANDBY LIQUID CONTROL (SLC), CONTROL ROD DRIVE (CRD) WATER (location SLC: unknown) (location CRD: unknown) Date Time Diff (hrs) Event/Action 3/11/2011 1446 0.00 seismic ground motions exceed setpoints 3/11/2011 1447 0.02 scram signal initiated, control rods inserted 3/11/2011 1447 0.02 EDGs start on loss of offsite power 3/11/2011 1527 0.68 first tsunami hits 3/11/2011 1535 0.82 second tsunami hits 3/11/2011 1541 0.92 loss of all AC power 3/11/2011 1541 0.92 loss of all DC power 3/11/2011 1541 0.92 SLC, CRD non-functional due to loss of AC power

Accident Timelines

2-8

UNIT 2: LOW PRESSURE SYSTEMS – CORE SPRAY (CS), RESIDUAL HEAT REMOVAL (RHR), RESIDUAL HEAT REMOVAL SEAWATER (RHRS) (location CS: RB basement, elev = O.P. -1.00m) (location RHR: RB basement, elev = O.P. -1.03m) (location RHRS: outdoors, elev = O.P. +4.00m) Date Time Diff (hrs) Event/Action 3/11/2011 1446 0.00 seismic ground motions exceed setpoints 3/11/2011 1447 0.02 scram signal initiated, control rods inserted 3/11/2011 1447 0.02 EDGs start on loss of offsite power 3/11/2011 1527 0.68 first tsunami hits 3/11/2011 1535 0.82 second tsunami hits 3/11/2011 1541 0.92 loss of all AC power 3/11/2011 1541 0.92 loss of all DC power

3/11/2011 1541 0.92 CS, RHR, RHRS non-functional due to loss of AC power

UNIT 2: OTHER POTENTIAL COOLING SYSTEMS – MAKEUP WATER (MUWC), FIRE PROTECTION, SPENT FUEL POOL (SFP) COOLING (location MUWC: Turbine Building (TB) basement, elev = O.P. +1.90m) (location fire protection: unknown) (location SFP: RB 3rd floor, elev = O.P. +26.90m) Date Time Diff (hrs) Event/Action 3/11/2011 1446 0.00 seismic ground motions exceed setpoints 3/11/2011 1447 0.02 scram signal initiated, control rods inserted 3/11/2011 1447 0.02 EDGs start on loss of offsite power 3/11/2011 1527 0.68 first tsunami hits 3/11/2011 1535 0.82 second tsunami hits 3/11/2011 1541 0.92 loss of all AC power 3/11/2011 1541 0.92 loss of all DC power

3/11/2011 1541 0.92 MUWC, SFP, motor-driven fire protection pump non-functional due to loss of AC power

Accident Timelines

2-9

Unit 3 UNIT 3: REACTOR CORE ISOLATION COOLING (RCIC) (location RCIC: RB basement, elev = O.P. -2.06m) Date Time Diff (hrs) Event/Action 3/11/2011 1446 0.00 seismic ground motions exceed setpoints 3/11/2011 1447 0.02 scram signal initiated, control rods inserted 3/11/2011 1448 0.03 EDGs started 3/11/2011 1505 0.32 RCIC manually started 3/11/2011 1525 0.65 RCIC tripped on high reactor water level 3/11/2011 1527 0.68 first tsunami hits 3/11/2011 1535 0.82 second tsunami hits 3/11/2011 1538 0.87 loss of all AC power 3/11/2011 1603 1.28 RCIC manually restarted 3/12/2011 1136 20.83 RCIC tripped 3/13/2011 0242 35.93 RCIC could not be restarted

UNIT 3: HIGH PRESSURE CORE INJECTION (HPCI) (location HPCI: RB basement, elev = O.P. -2.06m) Date Time Diff (hrs) Event/Action 3/11/2011 1446 0.00 seismic ground motions exceed setpoints 3/11/2011 1447 0.02 scram signal initiated, control rods inserted 3/11/2011 1448 0.03 EDGs start on loss of offsite power 3/11/2011 1527 0.68 first tsunami hits 3/11/2011 1535 0.82 second tsunami hits 3/11/2011 1538 0.87 loss of all AC power 3/11/2011 1603 1.28 HPCI prepared for startup 3/12/2011 1235 21.82 HPCI automatically started on low reactor water level 3/12/2011 1730 26.73 HPCI injection continues 3/13/2011 0242 35.93 HPCI tripped (all water injection sources lost) 3/13/2011 0242 35.93 HPCI could not be restarted due to depleted batteries

Accident Timelines

2-10

UNIT 3: OTHER HIGH PRESSURE SYSTEMS – STANDBY LIQUID CONTROL (SLC), CONTROL ROD DRIVE (CRD) WATER (location SLC: unknown) (location CRD: unknown) Date Time Diff (hrs) Event/Action 3/11/2011 1446 0.00 seismic ground motions exceed setpoints 3/11/2011 1447 0.02 scram signal initiated, control rods inserted 3/11/2011 1448 0.03 EDGs start on loss of offsite power 3/11/2011 1527 0.68 first tsunami hits 3/11/2011 1535 0.82 second tsunami hits 3/11/2011 1538 0.87 loss of all AC power 3/11/2011 1538 0.87 loss of all DC power 3/11/2011 1538 0.87 SLC, CRD non-functional due to loss of AC power

UNIT 3: LOW PRESSURE SYSTEMS – CORE SPRAY (CS), RESIDUAL HEAT REMOVAL (RHR), RESIDUAL HEAT REMOVAL SEAWATER (RHRS) (location CS: RB basement, elev = O.P. -1.00m) (location RHR: RB basement, elev = O.P. -1.03m) (location RHRS: outdoors, elev = O.P. +4.00m) Date Time Diff (hrs) Event/Action 3/11/2011 1446 0.00 seismic ground motions exceed setpoints 3/11/2011 1447 0.02 scram signal initiated, control rods inserted 3/11/2011 1448 0.03 EDGs start on loss of offsite power 3/11/2011 1527 0.68 first tsunami hits 3/11/2011 1535 0.82 second tsunami hits 3/11/2011 1538 0.87 loss of all AC power 3/11/2011 1538 0.87 loss of all DC power

3/11/2011 1538 0.87 CS, RHR, RHRS non-functional due to loss of AC power

Accident Timelines

2-11

UNIT 3: OTHER POTENTIAL COOLING SYSTEMS – MAKEUP WATER (MUWC), FIRE PROTECTION, SPENT FUEL POOL (SFP) COOLING (location MUWC: TB basement, elev = O.P. +2.42m) (location fire protection: unknown) (location SFP: RB 3rd floor, elev = O.P. +26.90m) Date Time Diff (hrs) Event/Action 3/11/2011 1446 0.00 seismic ground motions exceed setpoints 3/11/2011 1447 0.02 scram signal initiated, control rods inserted 3/11/2011 1448 0.03 EDGs start on loss of offsite power 3/11/2011 1527 0.68 first tsunami hits 3/11/2011 1535 0.82 second tsunami hits 3/11/2011 1538 0.87 loss of all AC power 3/11/2011 1538 0.87 loss of all DC power

3/11/2011 1538 0.87 Makeup water, motor-driven fire protection pump, containment vessel cooling non-functional due to loss of AC power

Accident Timelines

2-12

Table 2-1 Units 1 - 3 Power Source Unavailability Following the Earthquake/Tsunami

Unit Power Source Description of Unavailability Cause

Unit 1 Off-site power The receiving circuit breaker of the Ookuma Line 1L in the Unit 1 / 2 switchyard was damaged by the earthquake

Emergency diesel generator

Submergence of both emergency diesel generators due to the tsunami

6.9 kV high voltage AC power supply

Submergence of the 6.9 kV high voltage AC power supply panels due to the tsunami

480V low voltage AC power supply


Submergence of the 480V low voltage AC power supply panels due to the tsunami

125V DC power supply Submergence of the 125V DC power supply panels due to the tsunami

Unit 2 Off-site power source The receiving circuit breaker of the Ookuma Line 2L in the Unit 1 / 2 switchyard was damaged by the earthquake

The circuit breaker for the Ookuma Line 2L in the New Fukushima substation was damaged by the earthquake


One emergency diesel generator was submerged in water due to the tsunami and the power source panels for another, air cooled emergency diesel generator, was submerged due to the tsunami





Partial submergence of the 480V low voltage AC power supply panels due to the tsunami

125V DC power supply Submergence of the 125V DC power supply panels due to the tsunami

Unit 3 Off-site power Ookuma Line 4L was damaged by the earthquake between the plant switchyard and the off-site substation, Ookuma Line 3L was out of service for planned renovation work


Submergence of both emergency diesel generators due to the tsunami





Submergence of the 480V low voltage AC power supply panels due to the tsunami

125V DC power supply The DC power supply batteries were exhausted

Accident Timelines

2-13

Table 2-2 Status of Electrical Supply after Tsunami

Unit 1 Unit 2 Unit 3 Equipment

Name Status Equipment

Name Status Equipment

Name Status

EDG

EDG 1A* X EDG 2A* X EDG 3A* X EDG 1B* X EDG 2B air

cooled*** X EDG 3B* X

6.9

kV E

lect

rical

D

istr

ibut

ion Vi

tal M/C 1C** X M/C 2C* X M/C 3C* X

M/C 1D** X M/C 2D* X M/C 3D* X M/C 2E* X

Non

-Vita

l M/C 1A X M/C 2A X M/C 3A X M/C 1B X M/C 2B X M/C 3B X M/C 1S X M/C 2SA X M/C 3SA X

M/C 2SB X M/C 3SB X

480V

Pow

er C

ente

rs

(P/C

)

Vita

l P/C 1C X P/C 2C P/C 3C X P/C 1D X P/C 2D P/C 3D X

P/C 2E X

Non

-Vita

l

P/C 1A X P/C 2A P/C 3A X P/C 2A-1 X

P/C 1B X P/C 2B X P/C 3B P/C 1S X P/C 3SA

P/C 2SB X P/C 3SB

DC

Pow

er

125V

DC125V main bus A

X DC 125V P/C 2A

X DC 125V main bus

3A

O

DC125V main bus B

X DC 125V P/C 2B

X DC 125V main bus

3B

O

UH

S SW X RHR-S A X RHR-S A X RHR-S B X RHR-S B X

Status: “X” = Damaged, “O” = Available Not Damaged by Earthquake or Tsunami Damaged/Flooded by Tsunami Electric Distribution System Damaged/Flooded by Tsunami EDG – Emergency Diesel Generator * - Location, Turbine Building Basement M/C – Metal-Clad Switch Gear ** - Location, Turbine Building 1st Floor P/C – Power Center *** - Location, Shared Spent Fuel Pool 1st Floor UHS – Ultimate Heat Sink DC – Direct Current SW – Sea Water RHR – Residual Heat Removal

Accident Timelines

2-14

Excerpted from page IV-17 of Report of Japanese Government to the IAEA Ministerial Conference on Nuclear Safety - The Accident at TEPCO’s Fukushima Nuclear Power Stations, Prime Minister of Japan and His Cabinet [Online], June, 2011. http://www.kantei.go.jp/foreign/kan/topics/201106/ iaea_houkokusho_e.html. Partially modified by EPRI.

Figure 2-1 Flow Diagram of Cooling Systems, Unit 1, with ECCS Highlighted

http://www.kantei.go.jp/foreign/kan/topics/201106/%20iaea_houkokusho_e.html

http://www.kantei.go.jp/foreign/kan/topics/201106/%20iaea_houkokusho_e.html

Accident Timelines

2-15

Excerpted from page IV-17 of Report of Japanese Government to the IAEA Ministerial Conference on Nuclear Safety - The Accident at TEPCO’s Fukushima Nuclear Power Stations, Prime Minister of Japan and His Cabinet [Online], June, 2011. http://www.kantei.go.jp/foreign/kan/topics/201106/iaea_houkokusho_e.html. Partially modified by EPRI.

Figure 2-2 Flow Diagram of Cooling Systems, Units 2 and 3, with ECCS Highlighted

http://www.kantei.go.jp/foreign/kan/topics/201106/iaea_houkokusho_e.html

Accident Timelines

2-16

Excerpted from page IV-18 of Report of Japanese Government to the IAEA Ministerial Conference on Nuclear Safety - The Accident at TEPCO’s Fukushima Nuclear Power Stations, Prime Minister of Japan and His Cabinet [Online], June, 2011. http://www.kantei.go.jp/foreign/kan/topics/201106/iaea_houkokusho_e.html Partially modified by EPRI.

Figure 2-3 High Pressure Coolant Injection (HPCI) Flow Diagram, Units 1 – 3

Accident Timelines

2-17


Figure 2-4 Isolation Condenser (IC) Flow Diagram, Unit 1

Accident Timelines

2-18


Figure 2-5 Reactor Core Isolation Cooling (RCIC) Flow Diagram, Units 2 and 3

Accident Timelines

2-19

Excerpted from page IV-24 of Report of Japanese Government to the IAEA Ministerial Conference on Nuclear Safety - The Accident at TEPCO’s Fukushima Nuclear Power Stations, Prime Minister of Japan and His Cabinet [Online], June, 2011. http://www.kantei.go.jp/foreign/kan/topics/201106/iaea_houkokusho_e.html

Figure 2-6 Residual Heat Removal Flow Diagram, Units 2 and 3

Accident Timelines

2-20

Excerpted from page IV-23 of Report of Japanese Government to the IAEA Ministerial Conference on Nuclear Safety - The Accident at TEPCO’s Fukushima Nuclear Power Stations, Prime Minister of Japan and His Cabinet [Online], June, 2011. http://www.kantei.go.jp/foreign/kan/topics/201106/iaea_houkokusho_e.html

Figure 2-7 Reactor Shutdown Cooling (SHC) Flow Diagram, Unit 1

Accident Timelines

2-21


Figure 2-8 Main Steam Safety Relief Valve (MSRV) Flow Diagram, Unit 1

Accident Timelines

2-22


Figure 2-9 Main Steam Safety Relief Valve (MSRV) Flow Diagram, Units 2 and 3

Accident Timelines

2-23

Excerpted from page 13 of Attachment IV-3 of Report of Japanese Government to the IAEA Ministerial Conference on Nuclear Safety - The Accident at TEPCO’s Fukushima Nuclear Power Stations, Prime Minister of Japan and His Cabinet [Online], June, 2011. http://www.kantei.go.jp/foreign/kan/topics/201106/iaea_houkokusho_e.html

Figure 2-10 DC Power Supply System, Unit 1

Accident Timelines

2-24

Excerpted from page 14 of Attachment IV-3 of Report of Japanese Government to the IAEA Ministerial Conference on Nuclear Safety - The Accident at TEPCO’s Fukushima Nuclear Power Stations, Prime Minister of Japan and His Cabinet [Online], June, 2011. http://www.kantei.go.jp/foreign/kan/topics/201106/iaea_houkokusho_e.html

Figure 2-11 DC Power Supply System, Units 2 and 3

3-1

3 SEISMIC AND TSUNAMI DESIGN BASES

Because seismic and tsunami events were important factors in the accident at Fukushima Daiichi the following sections summarize the methods used to define the design bases for seismic and tsunami hazards at the plant. Criteria, methods, guidance, standards and regulations referred to in this report are those associated with Japan.

EPRI has accumulated significant experience with the Japanese nuclear industry in a number of technical areas such as maintenance, seismic, and water chemistry. This includes technical support provided to TEPCO, a member of EPRI’s nuclear sector since 2002, following the Niigataken-Chuetsu-Oki Earthquake that affected TEPCO’s Kashiwazaki-Kariwa nuclear power plant [14]. This experience provided background relevant to Japan seismic design basis and practice. Because EPRI had no experience with Japanese tsunami design basis or practice prior to the Fukushima Daiichi accident, the tsunami-related information used in this report derives from the references cited.

The design bases discussions include evaluations that have been performed since the original design and any associated upgrades that may have been implemented. Note that treatment of the design bases differed for earthquake and tsunami. Changes in the original seismic design basis resulted from a revised Japanese Regulatory Guide issued in 2006 [8]. A tsunami assessment method document issued by the Japan Society of Civil Engineers (JSCE) in 2002 [10] did not result in changes to the original design bases, but TEPCO did perform a voluntary reassessment of the tsunami design bases and implemented some plant design modifications.

Seismic Japanese regulators first issued general seismic design guidance in 1978, with subsequent revisions in 1981, 2001, and 2006 based in part on significant seismic events that occurred after 1978 [7, 8]. All plants, including Fukushima Daiichi, were required to be reviewed (and upgraded structurally if necessary) for conformance with this guidance.

Seismic and Tsunami Design Bases

3-2

The 2006 requirements for seismic event definition and qualification are specified in the Japan Nuclear Safety Commission (NSC) Regulatory Guide NSCRG: L-DS-I.02, entitled “Regulatory Guide for Reviewing Seismic Design of Nuclear Power Reactor Facilities” [8]. An English version of this document is available from the NSC website but readers are referred to the official version in Japanese “for all questions regarding meaning and phrasing.” The English version discusses basic policy and establishes the concept of “residual risks” which contemplates an earthquake stronger than the one specified in the design basis as follows:

“The seismic design shall be based on the principle to avoid any risks of serious radiological exposure to the public in the vicinity of the Facilities due to the external disturbance initiated by an earthquake, by appropriately formulating ‘the ground motion’ for the seismic design, which could be postulated appropriately to occur with a very low probability in the service period of the Facilities and could have serious affects to the Facilities.”

“Therefore, every effort should be made, at the design of the Facilities (in the basic design stage and subsequent stages), to minimize the ‘residual risks’ to the extent ‘as low as practically possible’ by appropriate attention to the possibility of ground motions exceeding the formulated design basis.”

The earthquakes that were defined for Fukushima Daiichi were upgraded after the issuance of NSCRG: L-DS-I.02. Three events postulated (1) Ss-1 inland crustal/interplate earthquakes, (2) Ss-2 oceanic intra-plate earthquakes, and (3) Ss-3 ground motion by unspecified source [9]. Each is shown in Figure 3-1 along with the original design earthquakes [9]. The new earthquakes are much stronger than the original earthquakes in the region with periods greater than 0.2 seconds (frequencies less than 5 Hz). The zero period accelerations are approximately 500 cm/s2, which is approximately 0.5 g. The probability of occurrence of these earthquakes, an approach to quantifying “residual risks”, has also been reported. The annual probability of exceedance for the response spectra was reported to be 10-4 to 10-6 [1].


3-3

Source: Excerpted with permission from Slide 15 of Fukushima Daiichi Nuclear Power Station, Fukushima Daini Nuclear Power Station, Seismic Ground Motion Due to Great East Japan Earthquake and Seismic Ground Motion Accounted for in Seismic Safety Assessments, Tokyo Electric Power Company.

Figure 3-1 Comparison of Original Seismic Spectra with Revised Seismic Spectra


3-4

Tsunami When the original licenses were issued for Fukushima Daiichi, Japanese regulatory guidance only stated that “(the effect of the) tsunami should be considered in design” [10]. There were no specific tsunami assessment numeric simulation methods available. For design purposes, therefore, the tsunami height was set at 3.1 meters above sea level, based on the observed wave height at Onahama port from the Chilean earthquake and tsunami of May 24, 1960.

Current Japanese regulatory guidance (in NSCRG L-DS-I.02 [8]) states that the “Safety functions of the Facilities shall not be significantly impaired by tsunami which could be reasonably postulated to hit in a very low probability in the service period of the Facilities.” More detailed technical guidance for nuclear power plants in Japan was provided in 2002 by the Japan Society of Civil Engineers (JSCE) [10] to effect implementation of this regulatory guidance. The methodology used historic records, if reliable, as the basis for the tsunami. The overall procedure from the English version of reference [10] obtained from the JSCE web site is as follows:

“(1) Tsunami source for the design tsunami

Among possible scenario tsunamis, the one causing the maximum water rise and fall at the site is selected as the “design tsunami.” The design water level is defined as the sum of the “design tsunami” and an appropriate tidal condition.

(2) A consideration policy with regard to the uncertainties of scenario tsunamis

In order to account for uncertainties regarding a tsunami source model, a large number of numerical calculations are carried out under various conditions within a reasonable range. Results from the parametric study are referred to as scenario tsunamis. This is referred to as a “parametric study”. Each results of the parametric study are termed as scenario tsunamis. For the model to the target site, the tsunami causing the greatest damage to the target site is selected among the scenario tsunamis.

(3) Method for verifying the design tsunami

The design tsunami is verified by using the following criteria.

The design tsunami height exceeds all recorded and calculated historical tsunami heights at the target site. In the vicinity of the target site, the envelope of the scenario tsunami heights exceeds all the recorded and calculated historical tsunami heights.

(4) Method for verifying the assessment procedure based on historical tsunamis

Before the above mentioned steps are carried out, a numerical calculation system is verified by performing numerical calculations on historical tsunamis.”

An important aspect of the 2002 JSCE guidance was that it did not consider as credible that a tsunami could be caused by ruptures across several fault segments in the vicinity of the plant. This was based on specific guidance provided in the 2002 JSCE document defining standard fault model parameters to use in establishing the design basis tsunami for nuclear power plants in Japan [10]. The JSCE guidance stated that combined fault segments did not need to be


3-5

considered for faults along the Japan Trench (the Japan Trench passes through the sea area off the coast of Japan that encompasses the region of Fukushima) [10, 16]. The March 11 earthquake occurred across numerous of the geological fault segments within the Japan Trench, resulting in a larger-than-expected tsunami [1, 4, 5, 6, 15].

It is also noted that the original and the revised guidance in the 2002 JSCE document addressed tsunami inundation height and that evaluation of loading conditions that could occur by the hydrodynamic effects (forces of moving water) of the tsunami was acknowledged but not specifically required [1, 10].

TEPCO applied the methods described in [10] considering tsunamis generated from eight different near-field sources off the coast of Japan. From this, it was determined that the wave height could reach 5.7m at Fukushima Daiichi. As these changes in criteria were voluntary, the licensing basis was not changed, although TEPCO made changes to assure that all vital seawater pump motors were installed higher than 5.7m. In conjunction with the revised Japan seismic Regulatory Guide [8] issued in 2006 TEPCO conducted a tsunami reevaluation using the methods in [10]. From this reevaluation, that incorporated updated submarine topography and tide level data, it was determined that the wave height could reach 6.1m at Fukushima Daiichi and additional plant actions such as sealing of pump motors were taken [17].

In 2006, TEPCO evaluated the probability of occurrence of a credible tsunami. The study estimated the probability of the coast near the Fukushima Daiichi plant experiencing a tsunami greater than 6m to be less than 10-2 in the following 50-year period [6]. This is equivalent to an annual probability of exceedance of more than 2 x10- 4.

In 2008, calculations by TEPCO to characterize a potential tsunami source without an established wave source model resulted in an estimated tsunami wave height of up to 10.2m and a resulting flood inundation height of over 15m at Fukushima Daiichi. Since a large earthquake had not occurred along the Japan Trench off the coast of Fukushima, a wave source from another location was applied off the coast of the Fukushima Daiichi plant to provide the “most strict” wave source for use in this estimate [17]. Another method that applied a wave source model of the Jogan tsunami in 869 A.D. resulted in an estimated wave height exceeding 9m [11].1 Because the 10.2m wave height estimate was from a wave source model not characteristic of the Japan Trench off the coast of Fukushima that had been analytically transposed to a location off the coast of the plant and the 9m height estimate was from a source model that had not yet been finalized neither of these estimated results were applied [11, 17].

In 2009, the Nuclear and Industrial Safety Agency (NISA) asked that operators take into account the Jogan earthquake for evaluating tsunami height “when new knowledge on the tsunami of the Jogan earthquake is obtained” [4]. The difficulties in establishing credible tsunami heights in the area from historical records are explained in more detail in [12]. Reference [17] discusses Jogan tsunami deposit surveys conducted by TEPCO in the Fukushima Prefecture. The survey results were reported in January 2011 and were inconsistent with the estimate using the Jogan tsunami wave source model used in earlier calculations; therefore TEPCO considered that it was necessary to conduct further investigation to determine the Jogan tsunami wave source [17].

1 The specific values of 9m and 15m are given in the cited reference. In the second report to the IAEA by the Japanese government, the calculated value is identified as “10m or higher” (chapter II of [4]).

4-1

4 MARCH 11 SEISMIC EVENT

The earthquake on March 11 occurred at 14:46 local time. The Magnitude 9 earthquake was the fourth largest in recorded world history. The epicenter was 111 miles (178 km) from the Fukushima Daiichi site. The peak ground acceleration in the horizontal direction was 0.56 g and in the vertical direction was 0.31 g as measured at Unit 2. Table 4-1 provides the maximum zero period accelerations measured at each plant [6]. Also presented are peak east-west spectral accelerations (all frequencies) as measured and as given within the design basis data (note that these values are scaled from plots in [6] and, as such, are approximate). All accelerations in this table are given in g’s (values in [6] are in Gals2). Ratios between measured and design basis values are also indicated. Ratios greater than 1.00 are highlighted.

Table 4-1 Maximum Measured Acceleration Values Compared to the Design Basis Acceleration Values

Measured (g) Max Design Basis (g) Ratio of Measured to Design Basis

Unit NS EW VERT Peak EW NS EW VERT Peak

EW NS EW VERT Peak EW

1 0.47 0.46 0.26 0.91 0.50 0.50 0.42 1.51 0.94 0.91 0.63 0.60

2 0.35 0.56 0.31 1.23 0.45 0.45 0.43 1.45 0.79 1.26 0.72 0.85

3 0.33 0.52 0.24 1.50 0.46 0.45 0.44 1.49 0.72 1.15 0.54 1.01

4 0.29 0.33 0.20 0.87 0.46 0.45 0.43 1.46 0.63 0.72 0.47 0.60

5 0.32 0.56 0.26 1.17 0.46 0.46 0.44 1.49 0.69 1.21 0.60 0.79

6 0.30 0.45 0.25 1.03 0.45 0.46 0.42 1.40 0.67 0.99 0.59 0.73

The units in operation at the time of the event automatically shutdown as a result of motions caused by the earthquake. Automatic shutdown of the units was set to occur when in-plant monitors reached horizontal accelerations of 0.14 g’s (or vertical accelerations of 0.10 g’s).

The response spectra at the sites are important indicators of the severity of the earthquake. The response spectra at Units 2, 3, and 5 are provided in Figure 4.1. Units 2, 3, and 5 had the largest spectral discrepancy (exceedance of the actual acceleration over the design acceleration) of the six units. Note that the exceedance of the actual spectra over the design spectra occurs primarily at frequencies between 2.5 and 5 Hz. (periods between 0.2 and 0.4 seconds). These are considered low frequencies and only a small amount of safety-related equipment that has natural frequencies in this range.

2 981 Gal = 1 g; 1 Gal = 0.01 m/s².

March 11 Seismic Event

4-2

The tsunami occurred approximately 41 minutes after the earthquake. The time elapsed between the seismic event and the first tsunami allowed clear indications of the response of the plants to the seismic event. Off-site AC power was also lost at all six units. Reports of the event indicate that the plants responded appropriately following the earthquake prior to the tsunami [6]. Control rods were inserted, diesel generators started, and all normal operator actions were being taken to respond to the automatic plant shutdown.

Figure 4-1 Measured Horizontal East-West Seismic Spectra Compared to the Design Basis Spectra for Units 2, 3, and 5

5-1

5 MARCH 11 TSUNAMI

At 15:27, 41 minutes after the seismic event, the first of a reported seven tsunami waves arrived at the site. By observing the height of water marks left on the buildings after the event TEPCO estimated that the maximum tsunami inundation height was 14 to 15 meters. The design basis height, as described in Section 3, was 6.1 meters.

The height of the tsunami exceeded the design basis because of the unexpected size of the seismic dislocation area, which was approximately 400 km (250 miles) north-south by 200 km (125 miles) east-west that involved rupture of multiple geological fault segments [1, 4, 15].

Units 1 to 4 were flooded to a depth of 4 to 5 meters above grade. There was extensive damage to site buildings and the turbine and reactor buildings were flooded. The intake structures at all six units were damaged by debris and flooding; and experienced electrical faults.

All AC power was lost at Units 1 to 5 when the diesel generator units and switchgear were flooded. DC power was lost at Units 1 and 2. The seawater intake structures were damaged. One of the diesel generators, an air-cooled diesel generator, remained functional at Unit 6.

The extent of flooding is shown in Figures 5-1 through 5-5 [3, 18].

March 11 Tsunami

5-2

Figure 5-1 Plan View of Site (Units 1 to 4 Only) Showing Flooded Regions Following the Tsunami

March 11 Tsunami

5-3

Figure 5-2 Section View (Approximate) through Unit 1 Showing Flood Depth

Source: Excerpted with permission from Result of the Investigation on Tsunami at Fukushima Daiichi Nuclear Power Station – Appendix 1, Press Release http://www.tepco.co.jp/en/press/corp-com/release/11040910-e.html, Tokyo Electric Power Company, April 9, 2011.

Figure 5-3 Photos at Points A and B in Figure 5-1

March 11 Tsunami

5-4

Source: Excerpted with permission from Slide 9 of IEEE Nuclear Power Engineering Committee Meeting - Fukushima Nuclear Accident Interim Report: Effects of the Earthquake and Tsunami on the Fukushima Daiichi and Daini Nuclear Power Stations, especially on electric and I&C systems and equipments, Kawano, Akira, Tokyo Electric Power Company, July 27, 2011.

Figure 5-4 Photos at Point C Showing Flooding Progression

March 11 Tsunami

5-5

Source: Excerpted with permission from Slide 8 of IEEE Nuclear Power Engineering Committee Meeting - Fukushima Nuclear Accident Interim Report: Effects of the Earthquake and Tsunami on the Fukushima Daiichi and Daini Nuclear Power Stations, especially on electric and I&C systems and equipments, Kawano, Akira, Tokyo Electric Power Company, July 27, 2011.

Figure 5-5 Photos Showing Seawall During the Tsunami

6-1

6 PLANT CAPABILITIES AND RESPONSE

The Fukushima Daiichi units had various structures, systems, and components (SSC) in place to respond to the events of March 11, 2011. For purposes here, these SSCs are grouped into operational, seismic, and tsunami capabilities.

Operational Capabilities and Response There were various systems designed to bring the reactors into – and maintain them at – safe shutdown conditions. Primary among these is the emergency core cooling systems (ECCSs), which is a composite of several high and low pressure systems with core cooling potential, intended to provide cooling during the decreasing pressure stages of a shutdown. There are additional systems which could have provided core cooling capabilities based on various valve lineups. However, as seen in the timeline descriptions in Chapter 2, all these systems were lost, most shortly after arrival of the second tsunami. The only exceptions to this near immediate loss were the IC system of Unit 1, the RCIC systems on Units 2 and 3, and the HPCI system on Unit 3. But even for these exceptions, the lack of electrical power eventually affected their ability to function. The time of failure of each of these systems relative to the initial seismic event and second tsunami are summarized in Table 6-1. Also shown in Table 6-1 is the point in time where lack of core cooling led to conditions believed to be sufficient to produce catastrophic failures, explosions and loss of primary containment.

Table 6-1 Timeline of Key Cooling Systems Failures

Unit System Hrs Action

All - 0.00 Earthquake

All - 0.82 Second tsunami wave

1 IC 11.03 Unit 1 IC lost

1 - 24.83 Unit 1 reactor building hydrogen explosion

2 RCIC 70.65 Unit 2 RCIC lost

2 - 87.23 Unit 2 loss of primary containment

3 RCIC 35.93 Unit 3 RCIC lost

3 HPCI 35.93 Unit 3 HPCI lost

3 - 68.25 Unit 3 reactor building hydrogen explosion

Plant Capabilities and Response

6-2

Seismic Capabilities and Response As indicated in Chapter 3, the seismic design bases established for all units were generally of such magnitudes and frequency that only small exceedances were noted with the actual event (see Chapter 4) even though the earthquake of March 11 was severe by historical standards.

The 41-minute period between the earthquake and first tsunami allowed clear indications of the plants’ responses to the seismic event and resulting loss of off-site power. Units 1, 2, and 3 were in operation, and – as designed – tripped due to the magnitude of the seismic event. Descriptions of the plants’ responses indicate that the plants responded appropriately following the earthquake (and prior to the tsunami) including control rod insertion, start of diesel generators, and operator actions [1, 6]. There were no indications via the control rooms (or anecdotally from plant personnel) of SSC failures that might have impeded the planned responses in each unit.

In many countries, plant seismic structural capabilities include collective conservatisms from known safety factors incorporated into design codes, damping factors, design redundancies (structural and operational), typical material margins (that is actual strengths above minimum required strengths), practical design considerations (for example off-the-shelf section strengths greater than minimum required), assumption of elastic response when plasticity will absorb energy and additional margins that may be incorporated in the design of specific structures. As such, appropriate responses as discussed above of the units to an actual earthquake that exceeded the seismic design bases would not be unexpected. Observations at the Kashiwazaki-Kariwa nuclear power plant in 2007 following the Niigataken-Chuetsu-Oki Earthquake were consistent with these seismic structural capabilities and plant response [14].

Tsunami Capabilities and Response As discussed in Chapter 3, the plant siting and subsequent design were originally based on a tsunami height of O.P. + 3.1m. The water intake structure was based on a maximum water level of O.P. + 4m and the Units 1 to 4 grade was set at O.P. + 10m. The grade for Units 5 and 6 was set at O.P. + 13m. The maximum seawall height was also set at O.P. + 10m.3 In subsequent tsunami reevaluations by TEPCO it was concluded that a tsunami height of O.P. + 5.7m and later a tsunami height of O.P. + 6.1m were possible that resulted in plant changes such as increasing the elevation of and sealing of seawater pump motors. The actual tsunami heights exceeded all of these values.

Because of the actual tsunami heights, significant plant flooding (Units 1 to 4) occurred in and around all major structures causing damage primarily by direct immersion in water. Additional damage was reported as a result of impact forces both from debris and hydrodynamic effects of the tsunami.

3 The seawall (breakwater) height varies between O.P. + 5.5m and O.P. + 10m. The seawall (breakwater) is not a full barrier but is one designed to (1) minimize wave action in the harbor, and (2) allow seaway access to the site dock between Units 1 and 5 – see the aerial photo in Figure 5-4 for the overall seawall layout; Figure 1-1 shows a docked ship.

7-1

7 CAUSAL ANALYSIS

Plant Protection As stated in Section 3, the original design basis tsunami height was established as O.P. + 3.1m based on a historical tsunami that had occurred in the region. Later assessments were done based on consideration of multiple historic tsunamis; these reviews resulted in a predicted tsunami height of O.P. + 5.7m and later O.P. + 6.1m. Changes to some of the sea water pump motor elevations and pump motor sealing were instituted as a result of these evaluations. Because the original siting for Units 1 to 4 was at a grade of O.P. + 10m, no additional major protection measures were made.

Thus, the primary plant protection for the effects of a tsunami involved the siting of the plant. That is, with all important systems and structures above the design basis tsunami height of 3.1m (or the adjusted height of 6.1m), the grade elevation difference with the tsunami height provided the protection. Although the majority of diesel generators were located in the basements of the turbine buildings – 7m to 8m below the plant grade – the grade elevation of the turbine buildings was above the design basis tsunami height.

Actual Event As discussed in Section 4, the initiating seismic event involved multiple ruptures of seismic sources over an area of about 250 miles x 125 miles. The earthquake was very significant (magnitude 9), considered the fourth largest in recorded world history. Although the earthquake did not cause significant structural or operational damage to Fukushima Daiichi, the event did cause major infrastructure damage to areas around the plant. The offsite damages led to loss of offsite power.

As discussed in Section 5, the earthquake caused a series of tsunamis, the largest of which arrived at Fukushima Daiichi approximately 41 minutes after the earthquake, reaching a wave height of approximately 50 ft (15 m). The associated volume of water – and the related hydrodynamic forces – caused extensive and deep flooding in and around all major structures of operating Units 1 through 3.

Comparison of Design Basis and Actual Event The design basis seismic definitions were – in magnitude and frequency content – not significantly different than the actual seismic event (see Section 4). However, the nature of the seismic event (that is, occurring across a large area and involving multiple ruptures of seismic fault segments) was not incorporated into the design basis tsunami definition.

Causal Analysis

7-2

The initial design basis tsunami height of 3.1m was based on typical tsunami prediction methods in use at the time of initial plant design. The later height of 6.1m was consistent with the Japanese industry accepted methods established in 2002. However, this height was well below the actual tsunami maximum height of approximately 15m, itself about 5m above plant grade (see Chapter 5). Based on the operational responses cited in Chapter 6, this height differential – along with the impact forces of the water (hydrodynamic effects) and debris – was the dominant cause for eventual loss of all practical cooling paths, damage to the reactor cores and uncontrolled release of radioactive materials to the environment.

Causal Analysis To assist in integrating and assessing information from the Fukushima Daiichi accident, an events and causal factors analysis approach was used to structure event and condition relationships [13]. This approach helps identify the underlying technical causal factors. The causal chart in Figure 7-1 displays technical cause and effect for the Fukushima Daiichi accident generally proceeding from left to right. Although there may be unit-specific differences, the causal chart shown here was constructed at a level such that the three units are depicted collectively. Based on the preceding discussions and assessments, the primary cause is that accepted Japanese nuclear industry guidance specified that combinations of geological fault segments in the vicinity of the plant need not be considered in establishing the design basis for tsunami height and that beyond-design-basis tsunami protection or mitigation capabilities were not able to address the actual event.

Causal Analysis

7-3

Figure 7-1 Causal Chart

8-1

8 CONCLUSIONS

A technical analysis of the Fukushima Daiichi March 2011 accident was conducted to determine the fundamental cause for the loss of substantial systems needed to maintain reactor cooling. The loss of these systems resulted in core damage to reactors at the site and uncontrolled release of radioactive materials to the environment from the site.

The record of the accident contained in the following documents was used as the basic source of information describing what occurred during the accident:

• Report of the Japanese Government to the IAEA Ministerial Conference on Nuclear Safety - The Accident at TEPCO’s Fukushima Nuclear Power Stations (Japanese Government)

• Examination of the Accident of the Tokyo Electric Power Co., Inc.'s Fukushima Daiichi Nuclear Power Station and Proposal of Countermeasures (Japan Nuclear Technology Institute)

• Special Report on the Nuclear Accident at the Fukushima Daiichi Nuclear Power Station (Institute of Nuclear Power Operations)

From this information and review of the capabilities needed to provide core cooling, it is clear that essentially all plant equipment needed to support core cooling was damaged by the initial effects of the tsunami event.

Other factors outside of the initial effects of the tsunami may have contributed to the extreme challenges encountered in attempts to sustain and/or reestablish cooling; however, the focus of this analysis was on the cause of the loss of the safety systems that would normally be used to maintain the integrity of the core. The loss of those safety systems was a result of a tsunami that exceeded the design basis of the plants.

The tsunami protection strategy for the plant consisted of locating critical equipment, such as vital seawater pump motors, above the elevation of the assessed tsunami height. The assessment of the tsunami height was based on the 2002 Tsunami Assessment Method for Nuclear Power Plants in Japan, the accepted methodology by the Japanese industry. Following the guidance of this methodology, offshore fault segments were not combined in the tsunami assessment. During the earthquake event numerous fault segments acted in combination, and thus the actual tsunami caused by the earthquake significantly exceeded the tsunami assessment for the plant.

Fundamentally, following the accepted tsunami assessment technical guidance in Japan resulted in under-prediction of the size of the tsunami. As a result, the plant tsunami protection strategy was not adequate and beyond-design-basis tsunami protection adequate to mitigate the effects of the tsunami that occurred was not available.

9-1

9 REFERENCES

1. Report of the Japanese Government to the IAEA Ministerial Conference on Nuclear Safety - The Accident at TEPCO’s Fukushima Nuclear Power Stations, Nuclear Emergency Response Headquarters, Government of Japan, June, 2011.

2. Recommendations for Enhancing Reactor Safety in the 21st Century, The Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, U.S. Nuclear Regulatory Commission, July 12, 2011.

3. IEEE Nuclear Power Engineering Committee Meeting - Fukushima Nuclear Accident Interim Report: Effects of the Earthquake and Tsunami on the Fukushima Daiichi and Daini Nuclear Power Stations, especially on electric and I&C systems and equipments, Kawano, Akira, Tokyo Electric Power Company, July 27, 2011.

4. Additional Report of the Japanese Government to the IAEA - The Accident at TEPCO’s Fukushima Nuclear Power Stations (Second Report), Nuclear Emergency Response Headquarters, Government of Japan, September, 2011.

5. Examination of the Accident of the Tokyo Electric Power Co., Inc.'s Fukushima Daiichi Nuclear Power Station and Proposal of Countermeasures, Japan Nuclear Technology Institute (JANTI), October, 2011 (DRAFT).

6. Special Report on the Nuclear Accident at the Fukushima Daiichi Nuclear Power Station, No. INPO 11-005, Revision 0, November 2011.

7. Seismic Safety Evaluation in Reactor Establishment Permit, Report JNESeL-SE-005, Japan Nuclear Energy Safety Organization (JNES), 3/15/11.

8. Regulatory Guide for Reviewing Seismic Design of Nuclear Power Reactor Facilities, Document No. NSCRG: L-DS-I.02, September, 2006, (Japan) Nuclear Safety Commission (NSC).

9. Fukushima Daiichi Nuclear Power Station, Fukushima Daini Nuclear Power Station, Seismic Ground Motion Due to Great East Japan Earthquake and Seismic Ground Motion Accounted for in Seismic Safety Assessments, Tokyo Electric Power Company.

10. Tsunami Assessment Method for Nuclear Power Plants in Japan, The Tsunami Evaluation Subcommittee, The Nuclear Civil Engineering Committee, Japan Society of Civil Engineers (JSCE), 2002 and 2006.

11. Executive Summary of the Interim Report (Provisional), Investigation Committee on the Accidents at Fukushima Nuclear Power Stations of Tokyo Electric Power Company (ICANPS), December 26, 2011.

12. Interplate Megathrust Earthquakes and Tsunamis Along Japan Trench Offshore Northeast Japan, Koji Okumura, Hiroshima University, March 2011.

References

9-2

13. Events and Causal Factors Analysis, SCIE-DOE-01-TRAC-14-95, SCIENTECH, Inc., August 1995.

14. EPRI Independent Peer Review of the TEPCO Seismic Walkdown and Evaluation of the Kashiwazaki-Kariwa Nuclear Power Plants: A Study in Response to the July 16, 2007 NCO Earthquake. EPRI, Palo Alto, CA: 2007.

15. An Overview of the M9.0 Earthquake Off the Pacific Coast of Tohoku Region of Japan on March 11, 2011and Its Lessons, Omer Aydan, Tokai University, March 20, 2011.

16. International Conference on Advances in Nuclear Power Plants- Fukushima Accident: An Overview, Akira Omoto, University of Tokyo, May 3, 2011.

17. Fukushima Nuclear Accident Analysis Report (Interim Report), Tokyo Electric Power Company, December 2011.

18. Result of the Investigation on Tsunami at Fukushima Daiichi Nuclear Power Station – Appendix 1, Press Release http://www.tepco.co.jp/en/press/corp-com/release/11040910-e.html, Tokyo Electric Power Company, April 9, 2011.

http://www.tepco.co.jp/en/press/corp-com/release/11040910-e.html

http://www.tepco.co.jp/en/press/corp-com/release/11040910-e.html



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