Fukushima Daiichi Accident Technical Causal Factor Analysis

Transcription

1 Fukushima Daiichi Accident Technical Causal Factor Analysis 2012 TECHNICAL REPORT

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3 Fukushima Daiichi Accident Technical Causal Factor Analysis Final Report, March 2012 EPRI Project Manager K. Huffman This document does NOT meet the requirements of 10CFR50 Appendix B, 10CFR Part 21, ANSI N and/or the intent of ISO-9001 (1994) ELECTRIC POWER RESEARCH INSTITUTE 3420 Hillview Avenue, Palo Alto, California PO Box 10412, Palo Alto, California USA

4 DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES THIS DOCUMENT WAS PREPARED BY THE ORGANIZATION(S) NAMED BELOW AS AN ACCOUNT OF WORK SPONSORED OR COSPONSORED BY THE ELECTRIC POWER RESEARCH INSTITUTE, INC. (EPRI). NEITHER EPRI, ANY MEMBER OF EPRI, ANY COSPONSOR, THE ORGANIZATION(S) BELOW, NOR ANY PERSON ACTING ON BEHALF OF ANY OF THEM: (A) MAKES ANY WARRANTY OR REPRESENTATION WHATSOEVER, EXPRESS OR IMPLIED, (I) WITH RESPECT TO THE USE OF ANY INFORMATION, APPARATUS, METHOD, PROCESS, OR SIMILAR ITEM DISCLOSED IN THIS DOCUMENT, INCLUDING MERCHANTABILITY AND FITNESS FOR A PARTICULAR PURPOSE, OR (II) THAT SUCH USE DOES NOT INFRINGE ON OR INTERFERE WITH PRIVATELY OWNED RIGHTS, INCLUDING ANY PARTY'S INTELLECTUAL PROPERTY, OR (III) THAT THIS DOCUMENT IS SUITABLE TO ANY PARTICULAR USER'S CIRCUMSTANCE; OR (B) ASSUMES RESPONSIBILITY FOR ANY DAMAGES OR OTHER LIABILITY WHATSOEVER (INCLUDING ANY CONSEQUENTIAL DAMAGES, EVEN IF EPRI OR ANY EPRI REPRESENTATIVE HAS BEEN ADVISED OF THE POSSIBILITY OF SUCH DAMAGES) RESULTING FROM YOUR SELECTION OR USE OF THIS DOCUMENT OR ANY INFORMATION, APPARATUS, METHOD, PROCESS, OR SIMILAR ITEM DISCLOSED IN THIS DOCUMENT. REFERENCE HEREIN TO ANY SPECIFIC COMMERCIAL PRODUCT, PROCESS, OR SERVICE BY ITS TRADE NAME, TRADEMARK, MANUFACTURER, OR OTHERWISE, DOES NOT NECESSARILY CONSTITUTE OR IMPLY ITS ENDORSEMENT, RECOMMENDATION, OR FAVORING BY EPRI. THE FOLLOWING ORGANIZATIONS PREPARED THIS REPORT: Electric Power Research Institute (EPRI) Lucius Pitkin, Inc. THE TECHNICAL CONTENTS OF THIS DOCUMENT WERE NOT PREPARED IN ACCORDANCE WITH THE EPRI NUCLEAR QUALITY ASSURANCE PROGRAM MANUAL THAT FULFILLS THE REQUIREMENTS OF 10 CFR 50, APPENDIX B AND 10 CFR PART 21, ANSI N AND/OR THE INTENT OF ISO-9001 (1994). USE OF THE CONTENTS OF THIS DOCUMENT IN NUCLEAR SAFETY OR NUCLEAR QUALITY APPLICATIONS REQUIRES ADDITIONAL ACTIONS BY USER PURSUANT TO THEIR INTERNAL PROCEDURES. NOTE For further information about EPRI, call the EPRI Customer Assistance Center at or askepri@epri.com. Electric Power Research Institute, EPRI, and TOGETHER SHAPING THE FUTURE OF ELECTRICITY are registered service marks of the Electric Power Research Institute, Inc. Copyright 2012 Electric Power Research Institute, Inc. All rights reserved.

5 ACKNOWLEDGMENTS The following organizations prepared this report: Electric Power Research Institute (EPRI) 1300 West W.T. Harris Blvd. Charlotte, NC Principal Investigator K. Huffman Lucius Pitkin Inc. 304 Hudson St. New York, NY Principal Investigators P. Bruck T. Esselman P. Streeter This report describes research sponsored by EPRI. This publication is a corporate document that should be cited in the literature in the following manner: Fukushima Daiichi Accident Technical Causal Factor Analysis. EPRI, Palo Alto, CA: iii

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7 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 v

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9 CONTENTS 1 INTRODUCTION Background ACCIDENT TIMELINES Unit Unit Unit SEISMIC AND TSUNAMI DESIGN BASES Seismic Tsunami MARCH 11 SEISMIC EVENT MARCH 11 TSUNAMI PLANT CAPABILITIES AND RESPONSE Operational Capabilities and Response Seismic Capabilities and Response Tsunami Capabilities and Response CAUSAL ANALYSIS Plant Protection Actual Event Comparison of Design Basis and Actual Event Causal Analysis CONCLUSIONS REFERENCES vii

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11 LIST OF FIGURES Figure 1-1 Arrangement of Fukushima Daiichi Units Figure 2-1 Flow Diagram of Cooling Systems, Unit 1, with ECCS Highlighted Figure 2-2 Flow Diagram of Cooling Systems, Units 2 and 3, with ECCS Highlighted Figure 2-3 High Pressure Coolant Injection (HPCI) Flow Diagram, Units Figure 2-4 Isolation Condenser (IC) Flow Diagram, Unit Figure 2-5 Reactor Core Isolation Cooling (RCIC) Flow Diagram, Units 2 and Figure 2-6 Residual Heat Removal Flow Diagram, Units 2 and Figure 2-7 Reactor Shutdown Cooling (SHC) Flow Diagram, Unit Figure 2-8 Main Steam Safety Relief Valve (MSRV) Flow Diagram, Unit Figure 2-9 Main Steam Safety Relief Valve (MSRV) Flow Diagram, Units 2 and Figure 2-10 DC Power Supply System, Unit Figure 2-11 DC Power Supply System, Units 2 and Figure 3-1 Comparison of Original Seismic Spectra with Revised Seismic Spectra Figure 4-1 Measured Horizontal East-West Seismic Spectra Compared to the Design Basis Spectra for Units 2, 3, and Figure 5-1 Plan View of Site (Units 1 to 4 Only) Showing Flooded Regions Following the Tsunami Figure 5-2 Section View (Approximate) through Unit 1 Showing Flood Depth Figure 5-3 Photos at Points A and B in Figure Figure 5-4 Photos at Point C Showing Flooding Progression Figure 5-5 Photos Showing Seawall During the Tsunami Figure 7-1 Causal Chart ix

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13 LIST OF TABLES Table 1-1 Comparison of Units 1 to Table 2-1 Units 1-3 Power Source Unavailability Following the Earthquake/Tsunami Table 2-2 Status of Electrical Supply after Tsunami Table 4-1 Maximum Measured Acceleration Values Compared to the Design Basis Acceleration Values Table 6-1 Timeline of Key Cooling Systems Failures xi

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15 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 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* BWR-3, Mark I IC, HPCI BWR-4, Mark I RCIC, HPCI BWR-4, Mark I RCIC, HPCI BWR-4, Mark I RCIC, HPCI BWR-4, Mark I RCIC, HPCI BWR-5, Mark II RCIC, HPCS * IC: Isolation Condenser HPCI: High Pressure Coolant Injection RCIC: Reactor Core Isolation Cooling HPCS: High Pressure Core Spray 1-1

16 Introduction 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 1-2

17 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 m. 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. 2-3

18 Accident Timelines Unit 1 UNIT 1: ISOLATION CONDENSER (IC) (location: Reactor Building (RB) 4 th floor, elev = O.P m) Date Time Diff (hrs) Event/Action 3/11/ seismic ground motions exceed setpoints 3/11/ scram signal initiated, control rods inserted 3/11/ Emergency diesel generators (EDGs) start on loss of offsite power 3/11/ IC's start automatically 3/11/ IC's manually removed from service 3/11/ IC(A) restarted 3/11/ IC(A) stopped 3/11/ IC(A) restarted 3/11/ IC(A) stopped 3/11/ first tsunami hits 3/11/ IC(A) restarted 3/11/ IC(A) stopped 3/11/ second tsunami hits 3/11/ loss of all AC power 3/11/ loss of all DC power 3/11/ valve status of IC lost on control panel 3/11/ partial DC power returned, IC(A) placed in service (steam observed from condenser) 3/11/ IC(A) removed from service 3/11/ IC(A) placed in service (steam observed) 3/12/ IC(A) stopped (diesel-driven fire protection pump failed) UNIT 1: HIGH PRESSURE CORE INJECTION (HPCI) (location: RB basement, elev = O.P m) Date Time Diff (hrs) Event/Action 3/11/ seismic ground motions exceed setpoints 3/11/ scram signal initiated, control rods inserted 3/11/ EDGs start on loss of offsite power 3/11/ first tsunami hits 3/11/ second tsunami hits 3/11/ loss of all AC power 3/11/ loss of all DC power 3/11/ HPCI determined to be non-functional 2-4

19 Accident Timelines 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/ seismic ground motions exceed setpoints 3/11/ scram signal initiated, control rods inserted 3/11/ EDGs start on loss of offsite power 3/11/ first tsunami hits 3/11/ second tsunami hits 3/11/ loss of all AC power 3/11/ loss of all DC power 3/11/ 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 m) (location CCSW: outdoors, elev = O.P m) (location MUWC: TB basement, elev = O.P m) (location fire protection: unknown) (location SFP: RB 3 rd floor, elev = O.P m) Date Time Diff (hrs) Event/Action 3/11/ seismic ground motions exceed setpoints 3/11/ scram signal initiated, control rods inserted 3/11/ EDGs start on loss of offsite power 3/11/ first tsunami hits 3/11/ second tsunami hits 3/11/ loss of all AC power 3/11/ loss of all DC power 3/11/ CCS, CCSW, MUWC, SFP, motor-driven fire protection pump non-functional due to loss of AC power 2-5

20 Accident Timelines Unit 2 UNIT 2: REACTOR CORE ISOLATION COOLING (RCIC) (location RCIC: RB basement, elev = O.P m) Date Time Diff (hrs) Event/Action 3/11/ seismic ground motions exceed setpoints 3/11/ scram signal initiated, control rods inserted 3/11/ EDGs start on loss of offsite power 3/11/ RCIC started by operators 3/11/ RCIC tripped due to high reactor water level 3/11/ RCIC started by operators 3/11/ first tsunami hits 3/11/ RCIC tripped due to high reactor water level 3/11/ second tsunami hits 3/11/ RCIC started by operators 3/11/ loss of all AC power 3/11/ loss of all DC power 3/11/ RCIC operational status unknown 3/11/ RCIC operational status still unknown; no reactor water level indication 3/12/ RCIC apparently still running based on walkdown (sounds only) 3/12/ RCIC apparently still running based on RCIC discharge pressure being high 3/12/ RCIC water source changed from condensate storage tank to suppression chamber 3/12/ RCIC still operating 3/13/ RCIC still operating 3/13/ RCIC still operating 3/14/ RCIC assumed failed based on lowered reactor water level 2-6

21 Accident Timelines UNIT 2: HIGH PRESSURE CORE INJECTION (HPCI) (location HPCI: RB basement, elev = O.P m) Date Time Diff (hrs) Event/Action 3/11/ seismic ground motions exceed setpoints 3/11/ scram signal initiated, control rods inserted 3/11/ EDGs start on loss of offsite power 3/11/ first tsunami hits 3/11/ second tsunami hits 3/11/ loss of all AC power 3/11/ loss of all DC power 3/11/ 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/ seismic ground motions exceed setpoints 3/11/ scram signal initiated, control rods inserted 3/11/ EDGs start on loss of offsite power 3/11/ first tsunami hits 3/11/ second tsunami hits 3/11/ loss of all AC power 3/11/ loss of all DC power 3/11/ SLC, CRD non-functional due to loss of AC power 2-7

22 Accident Timelines UNIT 2: LOW PRESSURE SYSTEMS CORE SPRAY (CS), RESIDUAL HEAT REMOVAL (RHR), RESIDUAL HEAT REMOVAL SEAWATER (RHRS) (location CS: RB basement, elev = O.P m) (location RHR: RB basement, elev = O.P m) (location RHRS: outdoors, elev = O.P m) Date Time Diff (hrs) Event/Action 3/11/ seismic ground motions exceed setpoints 3/11/ scram signal initiated, control rods inserted 3/11/ EDGs start on loss of offsite power 3/11/ first tsunami hits 3/11/ second tsunami hits 3/11/ loss of all AC power 3/11/ loss of all DC power 3/11/ 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 m) (location fire protection: unknown) (location SFP: RB 3 rd floor, elev = O.P m) Date Time Diff (hrs) Event/Action 3/11/ seismic ground motions exceed setpoints 3/11/ scram signal initiated, control rods inserted 3/11/ EDGs start on loss of offsite power 3/11/ first tsunami hits 3/11/ second tsunami hits 3/11/ loss of all AC power 3/11/ loss of all DC power 3/11/ MUWC, SFP, motor-driven fire protection pump nonfunctional due to loss of AC power 2-8

23 Accident Timelines Unit 3 UNIT 3: REACTOR CORE ISOLATION COOLING (RCIC) (location RCIC: RB basement, elev = O.P m) Date Time Diff (hrs) Event/Action 3/11/ seismic ground motions exceed setpoints 3/11/ scram signal initiated, control rods inserted 3/11/ EDGs started 3/11/ RCIC manually started 3/11/ RCIC tripped on high reactor water level 3/11/ first tsunami hits 3/11/ second tsunami hits 3/11/ loss of all AC power 3/11/ RCIC manually restarted 3/12/ RCIC tripped 3/13/ RCIC could not be restarted UNIT 3: HIGH PRESSURE CORE INJECTION (HPCI) (location HPCI: RB basement, elev = O.P m) Date Time Diff (hrs) Event/Action 3/11/ seismic ground motions exceed setpoints 3/11/ scram signal initiated, control rods inserted 3/11/ EDGs start on loss of offsite power 3/11/ first tsunami hits 3/11/ second tsunami hits 3/11/ loss of all AC power 3/11/ HPCI prepared for startup 3/12/ HPCI automatically started on low reactor water level 3/12/ HPCI injection continues 3/13/ HPCI tripped (all water injection sources lost) 3/13/ HPCI could not be restarted due to depleted batteries 2-9

24 Accident Timelines 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/ seismic ground motions exceed setpoints 3/11/ scram signal initiated, control rods inserted 3/11/ EDGs start on loss of offsite power 3/11/ first tsunami hits 3/11/ second tsunami hits 3/11/ loss of all AC power 3/11/ loss of all DC power 3/11/ 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 m) (location RHR: RB basement, elev = O.P m) (location RHRS: outdoors, elev = O.P m) Date Time Diff (hrs) Event/Action 3/11/ seismic ground motions exceed setpoints 3/11/ scram signal initiated, control rods inserted 3/11/ EDGs start on loss of offsite power 3/11/ first tsunami hits 3/11/ second tsunami hits 3/11/ loss of all AC power 3/11/ loss of all DC power 3/11/ CS, RHR, RHRS non-functional due to loss of AC power 2-10

25 Accident Timelines UNIT 3: OTHER POTENTIAL COOLING SYSTEMS MAKEUP WATER (MUWC), FIRE PROTECTION, SPENT FUEL POOL (SFP) COOLING (location MUWC: TB basement, elev = O.P m) (location fire protection: unknown) (location SFP: RB 3 rd floor, elev = O.P m) Date Time Diff (hrs) Event/Action 3/11/ seismic ground motions exceed setpoints 3/11/ scram signal initiated, control rods inserted 3/11/ EDGs start on loss of offsite power 3/11/ first tsunami hits 3/11/ second tsunami hits 3/11/ loss of all AC power 3/11/ loss of all DC power 3/11/ Makeup water, motor-driven fire protection pump, containment vessel cooling non-functional due to loss of AC power 2-11

26 Accident Timelines 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 6.9 kv high voltage AC power supply 480V low voltage AC power supply 125V DC power supply Submergence of both emergency diesel generators due to the tsunami Submergence of the 6.9 kv high voltage AC power supply panels due to the tsunami Submergence of the 6.9 kv high voltage AC power supply panels due to the tsunami Submergence of the 480V low voltage AC power supply panels due to the tsunami 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 Emergency diesel generator 6.9 kv high voltage AC power supply 480V low voltage AC power supply 125V DC power supply 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 Submergence of the 6.9 kv high voltage AC power supply panels due to the tsunami Submergence of the 6.9 kv high voltage AC power supply panels due to the tsunami Partial submergence of the 480V low voltage AC power supply panels due to the tsunami 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 Emergency diesel generator 6.9 kv high voltage AC power supply 480V low voltage AC power supply 125V DC power supply Submergence of both emergency diesel generators due to the tsunami Submergence of the 6.9 kv high voltage AC power supply panels due to the tsunami Submergence of the 6.9 kv high voltage AC power supply panels due to the tsunami Submergence of the 480V low voltage AC power supply panels due to the tsunami The DC power supply batteries were exhausted 2-12

27 Accident Timelines Table 2-2 Status of Electrical Supply after Tsunami EDG 6.9 kv Electrical Distribution 480V Power Centers (P/C) DC Power UHS 125V Non-Vital Vital Non-Vital Vital Unit 1 Unit 2 Unit 3 Equipment Name Status Equipment Name Status Equipment Name Status EDG 1A* X EDG 2A* X EDG 3A* X EDG 1B* X EDG 2B air EDG 3B* X X cooled*** 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 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 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 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 DC125V main bus A DC125V main bus B X X DC 125V P/C 2A DC 125V P/C 2B X X DC 125V main bus 3A DC 125V main bus 3B SW X RHR-S A X RHR-S A X RHR-S B X RHR-S B X O O 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 M/C Metal-Clad Switch Gear P/C Power Center UHS Ultimate Heat Sink DC Direct Current SW Sea Water RHR Residual Heat Removal * - Location, Turbine Building Basement ** - Location, Turbine Building 1 st Floor *** - Location, Shared Spent Fuel Pool 1 st Floor 2-13

28 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, iaea_houkokusho_e.html. Partially modified by EPRI. Figure 2-1 Flow Diagram of Cooling Systems, Unit 1, with ECCS Highlighted

29 Accident Timelines 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, Partially modified by EPRI. Figure 2-2 Flow Diagram of Cooling Systems, Units 2 and 3, with ECCS Highlighted 2-15

30 Accident Timelines 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, Partially modified by EPRI. Figure 2-3 High Pressure Coolant Injection (HPCI) Flow Diagram, Units

31 Accident Timelines Excerpted from page IV-19 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, Partially modified by EPRI. Figure 2-4 Isolation Condenser (IC) Flow Diagram, Unit

32 Accident Timelines Excerpted from page IV-20 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, Partially modified by EPRI. Figure 2-5 Reactor Core Isolation Cooling (RCIC) Flow Diagram, Units 2 and

33 Accident Timelines 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, Figure 2-6 Residual Heat Removal Flow Diagram, Units 2 and

34 Accident Timelines 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, Figure 2-7 Reactor Shutdown Cooling (SHC) Flow Diagram, Unit

35 Accident Timelines Excerpted from page IV-21 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, Partially modified by EPRI. Figure 2-8 Main Steam Safety Relief Valve (MSRV) Flow Diagram, Unit

36 Accident Timelines Excerpted from page IV-22 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, Partially modified by EPRI. Figure 2-9 Main Steam Safety Relief Valve (MSRV) Flow Diagram, Units 2 and

37 Accident Timelines 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, Figure 2-10 DC Power Supply System, Unit

38 Accident Timelines 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, Figure 2-11 DC Power Supply System, Units 2 and

39 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. 3-1

40 Seismic and Tsunami Design Bases 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/s 2, 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]. 3-2

41 Seismic and Tsunami Design Bases 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 3-3

42 Seismic and Tsunami Design Bases 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, 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 3-4

43 Seismic and Tsunami Design Bases 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]). 3-5

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45 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 Gals 2 ). 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 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 Gal = 1 g; 1 Gal = 0.01 m/s². 4-1

46 March 11 Seismic Event 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 4-2

47 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]. 5-1

48 March 11 Tsunami Figure 5-1 Plan View of Site (Units 1 to 4 Only) Showing Flooded Regions Following the Tsunami 5-2

49 March 11 Tsunami 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 Tokyo Electric Power Company, April 9, Figure 5-3 Photos at Points A and B in Figure

50 March 11 Tsunami 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, Figure 5-4 Photos at Point C Showing Flooding Progression 5-4

51 March 11 Tsunami 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, Figure 5-5 Photos Showing Seawall During the Tsunami 5-5

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53 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, 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 Earthquake All Second tsunami wave 1 IC Unit 1 IC lost Unit 1 reactor building hydrogen explosion 2 RCIC Unit 2 RCIC lost Unit 2 loss of primary containment 3 RCIC Unit 3 RCIC lost 3 HPCI Unit 3 HPCI lost Unit 3 reactor building hydrogen explosion 6-1