Text

APR1400-E-P-NR-14005-NP, Rev. 1, Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident.
	ML17114A520
Person / Time
Site:	05200046
Issue date:	03/31/2017
From:	; Korea Electric Power Corp, Korea Hydro & Nuclear Power Co, Ltd
To:	; Office of New Reactors
Shared Package
ML17114A517	List: ML17114A518; ML17114A520; ML17114A522; ML17114A524; ML17114A526;
References
	APR1400-E-P-NR-14005-NP, Rev 1
	Download: ML17114A520 (179)
	v • d • e

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident Revision 1 Non-Proprietary March 2017 Copyright 2017 Korea Electric Power Corporation &

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 REVISION HISTORY Sections Revision Date Description or Page(s)

December 0 All Initial Issue 2014 Added reference COL item related to developing Table 4-1, items shutdown risk process and procedures consistent 4.1, 4.2, 9.3 with NEI 12-06.

Modified content based on response to RAI 393-Section 5.1.2.1 8432_Q13 Rev.1 Modified content based on response to RAI 393-Section 5.1.2.2 8432_Q13 Rev.1 Removed information regarding shutdown mitigating Section 5.1.2.3 strategies based on response to RAI 393-8432_Q13 Rev.1 Section Modified statement and added table about room 5.1.2.3.1.1.2 temperatures

1. Typographical change based on response to RAI Section 297-8309_Q8 5.1.2.3.1.2.1 2. Modified statement and added table about room temperatures Water source sufficiency for phase 2 operation has Section been changed from 12 to 11 days based on response March 5.1.2.3.1.2.3 1 to RAI 517-8670_Q41 2017 Modified section and removed information regarding Section shutdown mitigating strategies based on response to 5.1.2.3.3 RAI 393-8432_Q13 Rev.1 Modified title/section and removed information Section regarding shutdown mitigating strategies based on 5.1.2.3.3.1 response to RAI 393-8432_Q13 Rev.1 and Q13 Rev.2 Modified content and removed information regarding Section shutdown mitigating strategies based on response to 5.1.2.3.3.2 RAI 393-8432_Q13 Rev.1 and Q13 Rev.2 Section Added information based on response to RAI 407-5.1.2.3.3.2 8447_Q29 Modified title and removed information regarding Section shutdown mitigating strategies based on response to 5.1.2.3.3.3 RAI 393-8432_Q13 Rev.1 Removed information regarding shutdown mitigating Section strategies based on response to RAI 393-8432_Q13 5.1.2.3.4 Rev.1 KEPCO & KHNP ii

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 REVISION HISTORY Sections Revision Date Description or Page(s)

Section Modified title and content based on response to RAI 5.1.2.3.4.2 393-8432_Q13 Rev.1 Deleted previous section 5.1.2.3.4.6 related to page 30 information regarding shutdown mitigating strategies Modified the following based on response to RAI 393-8432_Q13 Rev.1 1.Modified title

2. Adjusted section number due to deletion of Section previous section 5.1.2.3.4.3

3. Content of succeeding section has been deleted to this section

4. Added information about FLEX pumps and mobile GTGs in the ending part Deleted previous section 5.1.2.3.4.6 related to page 31 information regarding shutdown mitigating strategies March 1 Section Modified numerical data based on response to RAI 2017 5.1.2.4.1 517-8670_Q41 Modified the following based on response to RAI 517-8670_Q41:

Section

1. Modified numerical data 5.1.2.4.1.1

2. Added information.

3. Deleted sentence about non seismic piping.

Section Modified numerical data and some information based 5.1.2.4.1.2 on response to RAI 517-8670_Q41 Section Added information for the SFP Makeup FLEX Pumps 5.1.2.4.1.2 Section Modified numerical data and some information based 5.1.2.4.2.1 on response to RAI 517-8670_Q41 Section Modified numerical data and some information based 5.1.2.4.2.3 on response to RAI 517-8670_Q41 KEPCO & KHNP iii

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 REVISION HISTORY Sections Revision Date Description or Page(s)

1. Modified title and contents by removing information regarding shutdown mitigating strategies based on Sections 5.1.2.5 response to RAI 393-8432_Q13 Rev.1

2. Corrected typographical errors based on response 393 to RAI -8432_Q13 Rev.2 Modified title and content by removing information Sections regarding shutdown mitigating strategies based on 5.1.2.5.1 response to RAI 393-8432_Q13 Rev.1 and Rev.2 Section 5.1.2.5.2 Content has been revised and 2 subsections have Section been inserted as additional information based on 5.1.2.5.2.1 response to RAI 393-8432_Q13 Rev.1 Section 5.1.2.5.2.2 Modified the following based on response to RAI 393-8432_Q13 Rev.1 and Rev.2

1. Section number: 5.1.2.5.3 to 5.1.2.5.2.3 Section 2. Title and contents have been changed March 5.1.2.5.2.3 3. Superseded the added information in RAI 401-1 2017 8402_Q18

4. Superseded the added information in RAI 297-8309_Q1 Inserted section based on response to RAI 401-Section 8402_Q19, 401-8402_Q19 Rev.1 and 393-5.1.2.5.3 8432_Q13R2 Section Specified the type of current and provided additional 5.1.2.6.1.3 information for the emergency lightings.

Modified the following based on response to RAI Section response:

5.1.2.6.2 1. Added information on designing the RWT.

2. Modified numerical data Section Deleted portion of content based on response to RAI 5.1.2.6.3 393-8432_Q13 Rev.1 Section Modified numerical data based on response to RAI 5.1.3.2.1 517-8670_Q41 Section Modified numerical data based on response to RAI 5.1.3.2.2 517-8670_Q41 KEPCO & KHNP iv

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 REVISION HISTORY Sections Revision Date Description or Page(s)

Modified numerical data based on response to RAI Table 5-2 517-8670_Q41 Deleted previous Table 5-3 to remove information page 50 regarding shutdown mitigating strategies based on response to RAI 393-8432_Q13 Rev.1 Modified information based on response to RAI 517-Table 5-3 8670_Q41 Adjusted table number due to deletion of previous Table 5-4 table (Table 5-3)

Adjusted table number and removed information Table 5-5 regarding shutdown mitigating strategies based on response to RAI 393-8432_Q13 Rev.1

1. The following are changes based on response to RAI 393-8432_Q13 Rev.1:

March - Adjusted table number 1

2017 Table 5-6 - Modified contents related to Core Cooling and RCS Inventory, and Containment Integrity

2. Deleted some information based on response to RAI 393-8432_Q13 Rev.2

1. Adjusted table number

2. Deleted duration time in items 2.1, 2.2, 2.3 for initial phase, transition phase, and final phase respectively. In addition, portion of item 2.2 has been deleted

3. In item 5.0, the isolation valves needed to be opened by manual operation for FLEX strategies Table 5-7 have been specified

4. In item 5.1, the information regarding shutdown mitigating strategies has been deleted.

5. In item 2.2, transition phase extension has been changed from 12 days to 11 days

6. Modified some information based on response to RAI 393-8432_Q13 Rev.2 KEPCO & KHNP v

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 REVISION HISTORY Sections Revision Date Description or Page(s)

1. Adjusted table number

2. Modified statement in item 3.2.1.1 based on response to RAI 407-8447_Q25 Rev.1 Table 5-8

3. Modified statement in item 3.2.1.5 based on response to RAI 401-8402_Q14

4. Editorial correction

1. Adjusted table number

2. Added information that borated water is sourced Table 5-9 also from BAST

3. RWT water inventory minimum provision has been changed from 12 days to 11 days Adjusted table number due to deletion of previous Table 5-10 table (Table 5-3)

Deleted previous Figure 5-2 to removed information pages 97, 98 regarding shutdown mitigating strategies based on response to RAI 393-8432_Q13 Rev.1

1. Adjusted figure number

2. Replaced previous figure with revised figures March Figure 5-2 1 (Figure 5-2 and Figure 5-3) to removed information 2017 Figure 5-3 regarding shutdown mitigating strategies based on response to RAI 393-8432_Q13 Rev.1 Removed previous figures (Figure 5-4 and Figure 5-page 99 5) regarding shutdown mitigating strategies based on response to RAI 393-8432_Q13 Rev.1 Added information for overall description based on Section 6.1 response to RAI 297-8309_Q1 and 333-8397_Q9 Modified contents based on response to RAI 393-Section 6.2.2.1 8432_Q13 Rev.1

1. Modified content based on response to RAI 393-8432_Q13 Rev.1

2. Modified content based on response to RAI 297-Section 6.2.2.2 8309_Q4 (initial version, Rev.1, and Rev. 2)

3. Added information based on response to RAI 297-8309_Q1 Added information based on response to RAI 297-Section 6.2.3.1 8309_Q1 KEPCO & KHNP vi

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 REVISION HISTORY Sections Revision Date Description or Page(s)

1. Modified content based on response to RAI 297-8309_Q4 (initial version, Rev.1, and Rev. 2)

2. Deleted portion of content based on response to Section 6.2.3.2 RAI 393-8432_Q13 Rev.1

3. Modified content and added information based on response to RAI 297-8309_Q1 Modified numerical data based on response to RAI Section 6.2.3.2 517-8670_Q41 Modified numerical data based on response to RAI Section 6.2.4.1 517-8670_Q41 Modified numerical data based on response to RAI Section 6.2.5.2 517-8670_Q41 Modified content based on response to RAI 297-Section 6.2.5.2 8309_Q4 (initial version, Rev.1, and Rev. 2)

Replaced previous content based on response to RAI Section 6.2.9 March 333-8397_Q9 1

2017 Inserted section based on response to RAI 354-Section 6.2.10 8416_Q10 (initial version, and Rev.1)

Inserted section based on response to RAI 407-Section 6.2.11 8447_Q26 Rev.1

1. Updated DCD reference figures based on response to RAI 297-8309_Q1 Table 6-1

2. Inserted 4 but deleted 9 entries based on response to 401-8402_Q21 (initial version and Rev.1)

Table 6-2, 6-3, Added Tables 6-2, 6-3, 6-4, and 6-5 for additional 6-4, and 6-5 information Modified some information and added note based on Table 6-3 response to RAI 393-8432_Q13 Rev.2 Figure 6-5 Replaced Figure 6-5 to correct information Figure 6-7 Added Figure 6-7 for additional information KEPCO & KHNP vii

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 REVISION HISTORY Sections Revision Date Description or Page(s)

Section 8.0 Updated reference information Deleted portion of content regarding shutdown Appendix A - A.1 mitigating strategies based on response to RAI 393-8432_Q13 Rev.1 Appendix A - A.4 Deleted subsection title "A.4.1" as editorial correction.

Deleted content of succeeding section to previous page A2 A.4 regarding shutdown mitigating strategies based on response to RAI 393-8432_Q13 Rev.1 Deleted content of succeeding section to A.5.2.2 page A5 regarding shutdown mitigating strategies based on response to RAI 393-8432_Q13 Rev.1 Deleted Figure A-26, Figure A-27, and Figure A-28 page A19 based on response to RAI 393-8432_Q13 Rev.1 March Appendix B - Modified numerical data and added information 1

2017 Section B.2 based on response to RAI 517-8670_Q41 Appendix B - Modified numerical data based on response to RAI Section B.4 517-8670_Q41 Appendix B - Modified numerical data based on response to RAI Table B-1 517-8670_Q41 Appendix B - Modified numerical data based on response to RAI Table B-2 517-8670_Q41

1. Modified contents of Table B-3 based on response to RAI 393-8432_Q13 Rev.1, and 407-8447_Q28 Appendix B - 2. Modified Note 1 and deleted Note 2 of Table B-3 Table B-3 based on response to RAI 393-8432_Q13 Rev.1

3. Modified numerical data and some information based on response to RAI 517-8670_Q41 Appendix B - Modified numerical data in figure based on response Figure B-1 to RAI 517-8670_Q41 KEPCO & KHNP viii

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 This document was prepared for the design certification application to the U.S. Nuclear Regulatory Commission and contains technological information that constitutes intellectual property of Korea Hydro & Nuclear Power Co., Ltd. Copying, using, or distributing the information in this document in whole or in part is permitted only to the U.S. Nuclear Regulatory Commission and its contractors for the purpose of reviewing design certification application materials. Other uses are strictly prohibited without the written permission of Korea Electric Power Corporation and Korea Hydro &

Nuclear Power Co., Ltd.

KEPCO & KHNP ix

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 ABSTRACT The APR1400 mitigating strategies and design enhancements to meet Near-Term Task Force recommendations and U.S. NRC regulatory orders, relative to the Beyond Design Basis External Event at Fukushima Dai-ichi Nuclear Power Plant, are described in this technical report. In developing the mitigating strategies and design enhancements, the APR1400 design has considered the interim staff guidance and industry guidance. Accordingly, the APR1400 DCD reflects the lessons learned from the Fukushima BDBEE and its impact assessment, including identification of COL items. Any potential changes from proposed rulemaking to occur after the APR1400 DCD submittal have been excluded from this technical report.

Specifically, this technical report addresses the Near-Term Task Force Tier 1, 2, and 3 recommendations, and the multiple design changes considered for the APR1400 are supported by extensive analysis and calculations to address the Fukushima BDBEE. These enhancements have been incorporated into the APR1400 design; however, the operational and programmatic aspects of the BDBEE will be addressed by COL applicants prior to fuel load. Incorporation of the design enhancements, as described in this technical report, increases the APR1400 plant reliability and safety against BDBEE.

KEPCO & KHNP x

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 TABLE OF CONTENTS

1.0 INTRODUCTION

................................................................................................. 1 2.0 PURPOSE ............................................................................................................ 2 3.0 SCOPE ................................................................................................................ 3 4.0 REGULATORY RECOMMENDATIONS AND REQUIREMENTS .............................. 4 5.0 STRATEGIES TO ACTION ITEMS FROM FUKUSHIMA DAI-ICHI EVENTS ............................................................................................................ 23 5.1 Tier 1 Items .................................................................................................................................. 23 5.1.1 Recommendation 2.1 - Seismic and Flooding Re-Evaluation .................................................... 23 5.1.2 Recommendations 4.1 and 4.2 - SBO and FLEX ....................................................................... 23 5.1.3 Recommendation 7.1 - SFP Instrumentation .............................................................................. 38 5.1.4 Recommendation 8 - Emergency Response .............................................................................. 44 5.1.5 Recommendation 9.3 - Emergency Plan (only staffing and communications equipment portion in Tier 1) ......................................................................................................... 45 5.2 Tier 2 Items .................................................................................................................................. 45 5.2.1 Recommendation 7.2 - Safety-Related ac Electrical Power for the SFP Makeup System ......................................................................................................................................... 45 5.2.2 Recommendation 7.3 - Plant Technical Specifications ............................................................... 45 5.2.3 Recommendation 7.4 - Seismically Qualified Spent Fuel Pool Spray System ........................... 45 5.2.4 Recommendation 7.5 - Spent Fuel Pool Actions Related to Recommendations 7.1 through 7.4 ................................................................................................................................... 46 5.2.5 Recommendation 9.3 - Emergency Preparedness Regulatory Actions (remaining portions of Recommendation 9.3, except Emergency Response Data System -

ERDS capability addressed in Tier 3) .......................................................................................... 46 5.2.6 Recommendation 2.1 - Other External Events ........................................................................... 46 5.3 Tier 3 Items (and Other Items)..................................................................................................... 46 5.3.1 Recommendation 2.2 - Ten-Year Confirmation of Seismic and Flooding Hazards (dependent on Recommendation 2.1) ......................................................................................... 46 5.3.2 Recommendation 3 - Potential Enhancements to the Capability to Prevent or Mitigate Seismically-Induced Fires and Floods (Long-Term Evaluation) .................................... 46 5.3.3 Recommendation 5.2 - Reliable Hardened Vents for Other Containment Designs (Long-Term Evaluation) ............................................................................................................... 46 5.3.4 Recommendation 6 - Hydrogen Control and Mitigation inside Containment or in Other Buildings ............................................................................................................................ 47 5.3.5 Recommendations 9.1 and 9.2 - Emergency Preparedness (EP) Enhancements for Prolonged SBO and Multiunit Events .......................................................................................... 47 5.3.6 Recommendation 9.3 - ERDS Capability.................................................................................... 47 5.3.7 Recommendation 10 - Additional EP Topic for Prolonged SBO and Multiunit Events ............... 47 KEPCO & KHNP xi

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 5.3.8 Recommendation 11 - EP Topics for Decision-Making, Radiation Monitoring, and Public Education .......................................................................................................................... 47 5.3.9 Recommendation 12.1 - Reactor Oversight Process Modifications to Reflect the Recommended Defense-in-Depth Framework ............................................................................ 47 5.3.10 Recommendation 12.2 - Staff Training on Severe Accidents and Resident Inspector Training on SAMGs...................................................................................................................... 47 5.3.11 Additional Recommendations ...................................................................................................... 47 6.0 DESIGN FEATURES AND PROGRAMS TO ADDRESS BDBEE .......................... 100 6.1 Overall Description .................................................................................................................... 100 6.2 Specific Design Enhancements and Programs ......................................................................... 100 6.2.1 Beyond Design Basis Seismic and Flood Protection ................................................................ 100 6.2.2 Primary Side FLEX Pump(s) and Connections ......................................................................... 100 6.2.3 Spent Fuel Pool - Makeup Line and Spray Line Enhancements .............................................. 101 6.2.4 SFP Level Instrumentation ........................................................................................................ 102 6.2.5 AFWS Secondary Side FLEX Pump Connection ...................................................................... 103 6.2.6 Electric Power Supply System ................................................................................................... 104 6.2.7 Operational Program, Procedures, and Training ....................................................................... 105 6.2.8 Emergency Procedures ............................................................................................................. 105 6.2.9 Storage of FLEX Equipment ...................................................................................................... 105 6.2.10 Installed equipment and tanks utilized in the mitigation strategies............................................ 106 6.2.11 Connections for FLEX strategies ............................................................................................... 107

7.0 CONCLUSION

................................................................................................. 120

8.0 REFERENCES

................................................................................................. 121 KEPCO & KHNP xii

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 LIST OF TABLES Table 4-1 Post-Fukushima NRC Recommendations and Requirements (1 of 18)..................................... 5 Table 5-1 Sequence of Events for Core Cooling (Full-Power Operation) (1 of 2) .................................... 48 Table 5-2 Water Volume Source and Requirements for SG Feedwater................................................... 50 Table 5-3 FLEX Capability - Spent Fuel Pool Cooling Summary ............................................................ 51 Table 5-4 480 V and 4.16 kV Mobile GTG Electrical Load Summary List (in kW) ................................... 52 Table 5-5 Summary of Fuel Oil Demand (most limiting) ........................................................................... 53 Table 5-6 APR1400 FLEX Capability Summary (1 of 3) .......................................................................... 54 Table 5-7 Conformance with JLD-ISG-2012-01, Rev. 0 (1 of 5) ............................................................... 57 Table 5-8 Conformance with NEI 12-06, Rev. 0 (1 of 20) ......................................................................... 62 Table 5-9 Conformance with NEI 12-06, Rev. 0 - Tables D-1, D-2, and D-3 (1 of 4)............................... 82 Table 5-10 Conformance with NEI 12-02, Rev. 1 (1 of 11) ....................................................................... 86 Table 6-1 External Connection Components for BDBEE (1 of 2) ........................................................... 108 Table 6-2 List of Installed Safety Related Pumps and Valves for BDBEE (1 of 2) ................................. 110 Table 6-3 List of On-site FLEX Equipment for BDBEE ........................................................................... 112 Table 6-4 List of Off-site FLEX Equipment for BDBEE ........................................................................... 113 Table 6-5 List of Connection for FLEX strategies ................................................................................... 114 KEPCO & KHNP xiii

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 LIST OF FIGURES Figure 5-1 Timeline of the APR1400 FLEX Strategy for Full-Power Operation ....................................... 97 Figure 5-2 Containment Pressure for Full Power (RCP Seal LOCA) ....................................................... 98 Figure 5-3 Containment Temperature for Full Power (RCP Seal LOCA) ................................................. 99 Figure 6-1 Primary Side FLEX Pump Connection into the Safety Injection System .............................. 115 Figure 6-2 FLEX Pump Suction Source for SFP Makeup and Spray Line ............................................. 116 Figure 6-3 Connection for SFP Makeup and Spray Line ....................................................................... 116 Figure 6-4 Layout of SFP Makeup and SFP Spray Line Connections ................................................... 117 Figure 6-5 Water Supply System to the Secondary Side of SG ............................................................. 118 Figure 6-6 Fuel Oil Supply System to FLEX Pumps .............................................................................. 118 Figure 6-7 Flow Path for FLEX Connection to Deliver Water to Containment for ECSBS..................... 119 KEPCO & KHNP xiv

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 LIST OF APPENDICES APPENDIX. A SUPPORTING ANALYSIS RESULTS FOR THE OPERATIONAL STRATEGY FOR CORE COOLING .................................................. Ai APPENDIX. B SPENT FUEL POOL TIME TO BOIL AND MAKEUP ANALYSIS ............. Bi APPENDIX. C 480 V AND 4.16 KV MOBILE GTG ELECTRICAL LOADINGS ............... Ci KEPCO & KHNP xv

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 ACRONYMS AND ABBREVIATIONS AAC alternate alternating current ac alternating current ACP auxiliary charging pump AFWP auxiliary feedwater pump AFAS auxiliary feedwater actuation signal AFWS auxiliary feedwater system AFWST auxiliary feedwater storage tank AHU air handling unit ANPR Advance Notice of Proposed Rulemaking ANS American Nuclear Society ANSI American National Standards Institute AOP abnormal operating procedure AOV air-operated valve APR advanced power reactor BDB beyond design basis BDBEE beyond design basis external event BWR boiling water reactor CAV cumulative absolute velocity CCW component cooling water CEUS central and eastern United States CEUS-SSC CEUS seismic source characterization CFR Code of Federal Regulations CIV containment isolation valve COL combined license COLA combined license application CP construction permit DBA design basis accident dc direct current DCD design control document EA Enforcement Action EDG emergency diesel generator ECSBS emergency containment spray backup subsystem EDMG extensive damage mitigation guideline EFW emergency feedwater KEPCO & KHNP xvi

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 ELAP extended loss of ac power EOP emergency operating procedure EP emergency preparedness EPRI Electric Power Research Institute EPZ emergency planning zone ERDS emergency response data system ESW essential service water FEMA Federal Emergency Management Agency FIRS foundation input response spectra FLEX diverse and flexible coping strategies FSG FLEX support guideline GDC General Design Criteria GMRS ground motion response spectra GTG gas turbine generator GWR guided wave radar HPCI high-pressure coolant injection HVAC heating, ventilation and air conditioning Hx heat exchanger I&C instrumentation and control IEEE Institute of Electrical and Electronics Engineers IRWST in-containment refueling water storage tank ISG interim staff guidance KHNP Korea Hydro & Nuclear Power Company Ltd.

LOOP loss of offsite power LOLA loss of large area LTOP low-temperature overpressurizaton protection LOSCS loss of shutdown cooling system LUHS loss of normal access to ultimate heat sink MCC motor control center MCR main control room MSADV main steam atmospheric dump valve MSSV main steam safety valve MOV motor-operated valve MTC moderator temperature coefficient NA not applicable NCC natural circulation cooling KEPCO & KHNP xvii

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 NEI Nuclear Energy Institute NRC Nuclear Regulatory Commission NTTF Near-Term Task Force OSC operational support center PMF probable maximum flood PORV power-operated relief valve POSRV pilot-operated safety relief valve PSHA probabilistic seismic hazard analysis PWR pressurized water reactor PZR pressurizer RAI request for additional information RAT reserve auxiliary transformer RCIC reactor core isolation cooling RCS reactor coolant system RCP reactor coolant pump RG Regulatory Guide RFI Request for Information RHR residual heat removal RPS reactor protection system RSR remote shutdown room RV reactor vessel RWT raw water tank SAMG severe accident management guidelines SAT systematic approach to training SBO station blackout SCS shutdown cooling system SFP spent fuel pool SG steam generator SIP safety injection pump SIT safety injection tank SSC structure, system, or component TDAFWP turbine-driven auxiliary feedwater pump TDH total dynamic head TDR time domain reflectometry TSC technical support center UHS ultimate heat sink KEPCO & KHNP xviii

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 UPC ultimate pressure capacity UPS uninterruptible power system VBPSS vital bus power supply system KEPCO & KHNP xix

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1

1.0 INTRODUCTION

The March 11, 2011 earthquake and subsequent tsunami off the Pacific coast of Japan (Great Tohoku Earthquake) exceeded the seismic and tsunami design bases of the Fukushima Dai-ichi Nuclear Power Plant. The event resulted in major damage at the site. Subsequent evaluation by regulatory and industry experts resulted in insights that nuclear plants should have additional capability to withstand beyond-design-basis external events (BDBEEs). This capability would enhance protection against accidents resulting from natural phenomena, mitigate the consequences of such accidents, and enhance emergency preparedness. It reflects a diverse and flexible coping strategy to increase the defense-in-depth safety principle that has long been a foundation of the commercial nuclear power industry.

Korea Hydro & Nuclear Power Co., Ltd. (KHNP) has evaluated post-Fukushima insights in terms of its Advanced Power Reactor 1400 (APR1400) design. Sections 2.0 and 3.0 of this report summarize the purpose and scope, respectively. Section 4.0 summarizes applicable regulatory requirements and the potential effect on the APR1400 licensing documentation. Section 5.0 provides technical description of how the APR1400 addresses post-Fukushima insights. Section 6.0 identifies design features and their impact on the Design Control Document (DCD) and the combined license (COL) applicant. Section 7.0 provides the reports overall conclusions.

KEPCO & KHNP 1

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 2.0 PURPOSE The purpose of this report is to address requirements and guidance provided by the U.S. Nuclear Regulatory Commission (NRC) in a series of Commission papers (SECY), NRC Orders (provided via Enforcement Actions, EAs), and Interim Staff Guidance (ISG) after the Fukushima event. In addition, industry initiatives identified by the Nuclear Energy Institute (NEI) have been considered in the development of this report. The applicable requirements for the APR1400 design are summarized in Table 4-1.

KEPCO & KHNP 2

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 3.0 SCOPE This technical report addresses the lessons learned from the Fukushima event, which form an integral part of the APR1400 design. This document also identifies actions associated with insights from the Fukushima event. Any potential changes from proposed rulemaking that are scheduled to occur after the APR1400 is submitted for NRC design certification are outside of the scope of this report.

KEPCO & KHNP 3

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 4.0 REGULATORY RECOMMENDATIONS AND REQUIREMENTS This section addresses post-Fukushima NRC recommendations/requirements and actions by KHNP and/or the COL applicant for each recommendation and requirement. In this section, the following post-Fukushima NRC recommendations and requirements are addressed:

SECY 11-0093, Recommendations for Enhancing Reactor Safety in the 21st Century, The Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident (Near-Term Task Force (NTTF) Recommendations) (Reference 1)

SECY-11-0137, Prioritization of Recommended Actions to be Taken in Response to Fukushima Lessons Learned (Reference 2)

SECY-12-0025, Proposed Orders and Requests for Information in Response to Lessons Learned from Japans March 11, 2011, Great Tohoku Earthquake and Tsunami (Reference 3)

SECY-12-0095, Tier 3 Program Plans and 6-Month Status Update in Response to Lessons Learned from Japans March 11, 2011, Great Tohoku Earthquake and Subsequent Tsunami (Reference 4)

NRC Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Reference 5)

NRC Order EA-12-051, Reliable Spent Fuel Pool Instrumentation (Reference 6)

Table 4-1 provides a cross-reference to sections in this report where KHNP actions to address the recommendations specified in the above NRC documents are described. This table also lists those applicable DCD sections that address the recommendations and provides a brief summary of applicable actions for each recommendation. The post-Fukushima NRC recommendations that are not applicable to either the APR1400 DCD or COL applicant(s) are also identified in the table.

KEPCO & KHNP 4

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 4-1 Post-Fukushima NRC Recommendations and Requirements (1 of 18)

NTTF NRC Recommendations/Requirements in SECY-Applicable DCD Rec. 11-0093, SECY-11-0137, SECY-12-0025, SECY APR1400 Design COL Action Note Section No 0095, EA-12-049, EA-12-051 Tier 1 (Actions to be taken without delay) 2.1 NA NA COL 19.3(1) Request for Seismic Reevaluation information via a) Evaluate the potential impacts of the newly 50.54(f) letter.

released Central and Eastern United States Seismic Source Characterization (CEUS-SSC) model, with potential local and regional refinements as identified in the CEUS-SSC model, on the seismic hazard curves and the site-specific ground motion response spectra (GMRS)/foundation input response spectra (FIRS). For re-calculation of the probabilistic seismic hazard analysis (PSHA), please follow either the cumulative absolute velocity (CAV) filter or minimum magnitude specifications outlined in Attachment 1 to Seismic Enclosure 1 of the March 12, 2012 letter "Request for information pursuant to Title 10 of the Code of Federal Regulations 50.54(f) regarding Recommendations 2.1, 2.3, and 9.3, of the near-term task force review of insights from the Fukushima Dai-ichi accident." (ML12053A340).

b) In your response, please identify the method you selected from the above choices to perform the evaluation. Modify and submit the site-specific GMRS and FIRS changes, as necessary, given the evaluation performed in part (a) above. Provide the basis supporting your position.

KEPCO & KHNP 5

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 4-1 Post-Fukushima NRC Recommendations and Requirements (2 of 18)

NTTF NRC Recommendations/Requirements in SECY-Applicable DCD Rec. 11-0093, SECY-11-0137, SECY-12-0025, SECY APR1400 Design COL Action Note Section No 0095, EA-12-049, EA-12-051 2.1 Flooding Reevaluation NA NA COL 19.3(2) Request for information via

Perform a reevaluation of all appropriate external 50.54(f) letter.

flooding sources, including the effects from local intense precipitation on the site, probable maximum flood (PMF) on stream and rivers, storm surges, seiches, tsunami, and dam failures.

It is requested that the reevaluation apply present-day regulatory guidance and methodologies being used for ESP and COL reviews including current techniques, software, and methods used in present-day standard engineering practice to develop the flood hazard.

2.3 Seismic Walkdowns NA NA NA Request for information via

Perform seismic walkdowns in order to identify 50.54(f) letter.

and address plant specific degraded, non-conforming, or unanalyzed conditions and verify the adequacy of strategies, monitoring, and maintenance programs such that the nuclear power plant can respond to external events. The walkdown will verify current plant configuration with the current licensing basis, verify the adequacy of current strategies, maintenance plans, and identify degraded, non-conforming, or unanalyzed conditions.

KEPCO & KHNP 6

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 4-1 Post-Fukushima NRC Recommendations and Requirements (3 of 18)

NTTF NRC Recommendations/Requirements in SECY-Applicable DCD Rec. 11-0093, SECY-11-0137, SECY-12-0025, SECY APR1400 Design COL Action Note Section No 0095, EA-12-049, EA-12-051 2.3 Flooding Walkdowns NA NA NA Request for

Perform flood protection walkdowns using an information via NRC-endorsed walkdown methodology, 50.54(f) letter.

Identify and address plant-specific degraded, non-conforming, or unanalyzed conditions as well as cliff-edge effects through the corrective action program and consider these findings in the Recommendation 2.1 hazard evaluations, as appropriate,

Identify any other actions taken or planned to further enhance the site flood protection,

Verify the adequacy of programs, monitoring and maintenance for protection features, and,

Report to the NRC the results of the walkdowns and corrective actions taken or planned.

4.1 Station Blackout (SBO) See NTTF 19.3.2.3 COL 19.3(3),

Recommendations 19.3(4),19.3(5),

(NTTF Recommendations) Initiate rulemaking to 4.1 and 4.2 in and 19.3(17) revise 10 CFR 50.63 to require each operating and Subsection 5.1.2 new reactor licensee to (1) establish a minimum coping time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for a loss of all ac power, (2) establish the equipment, procedures, and training necessary to implement an extended loss of all ac coping time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for core and spent fuel pool cooling and for reactor coolant system and primary containment integrity as needed, and (3) preplan and prestage offsite resources to support uninterrupted core and spent fuel pool cooling, and reactor coolant system and containment integrity as needed, including the ability to deliver the equipment to the site in the time period allowed for extended coping, under conditions involving significant degradation of KEPCO & KHNP 7

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 4-1 Post-Fukushima NRC Recommendations and Requirements (4 of 18)

NTTF NRC Recommendations/Requirements in SECY-Applicable DCD Rec. 11-0093, SECY-11-0137, SECY-12-0025, SECY APR1400 Design COL Action Note Section No 0095, EA-12-049, EA-12-051 offsite transportation infrastructure associated with significant natural disasters.

4.2 Mitigation Strategies for Beyond-Design-Basis See NTTF 19.3.2.3 COL 19.3(3),

External Events Recommendation 19.3(4), 19.3(5),

(EA-12-049) 4.2 in Subsection and 19.3(17) 5.1.2 (1) Licensees shall develop, implement and maintain guidance and strategies to maintain or restore core cooling, containment and SFP cooling capabilities following a beyond-design-basis external event.

(2) These strategies must be capable of mitigating a simultaneous loss of all alternating current (ac) power and loss of normal access to the ultimate heat sink and have adequate capacity to address challenges to core cooling, containment, and SFP cooling capabilities at all units on a site subject to this Order.

(3) Licensee must provide reasonable protection for the associated equipment from external events.

Such protection must demonstrate that there is adequate capacity to address challenges to core cooling, containment, and SFP cooling capabilities at all units on a site subject to this order.

(4) Licensee must be capable of implementing the strategies in all modes.

(5) Full compliance shall include procedures, guidance, training, and acquisition, staging, or installation of equipment needed for the strategies.

KEPCO & KHNP 8

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 4-1 Post-Fukushima NRC Recommendations and Requirements (5 of 18)

NTTF NRC Recommendations/Requirements in SECY-Applicable DCD Rec. 11-0093, SECY-11-0137, SECY-12-0025, SECY APR1400 Design COL Action Note Section No 0095, EA-12-049, EA-12-051 5.1 NA NA NA Reliable Hardened Vents for Mark I and Mark II containments Boiling-Water Reactor (BWR) Mark I and Mark II containments shall have a reliable hardened vent to remove decay heat and maintain control of containment pressure within acceptable limits following events that result in the loss of active containment heat removal capability or prolonged Station Blackout (SBO). The hardened vent system shall be accessible and operable under a range of plant conditions, including a prolonged SBO and inadequate containment cooling.

7.1 See NTTF 19.3.2.4 COL 19.3(6)

SFP Instrumentation Recommendation (EA-12-051 to COL Holder) Item 7.1 in Licensee requires reliable indication of the water level Subsection 5.1.3 in associate spent fuel storage capable of supporting identification of the following pool water level conditions by trained personnel: (1) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement makeup water addition should no longer be deferred.

1. The spent fuel pool level instrumentation shall include the following design features:

KEPCO & KHNP 9

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 4-1 Post-Fukushima NRC Recommendations and Requirements (6 of 18)

NTTF NRC Recommendations/Requirements in SECY-Applicable DCD Rec. 11-0093, SECY-11-0137, SECY-12-0025, SECY APR1400 Design COL Action Note Section No 0095, EA-12-049, EA-12-051

1.1 Arrangement

The spent fuel pool level instrument channels shall be arranged in a manner that provides reasonable protection of the level indication function against missiles that may result from damage to the structure over the spent fuel pool. This protection may be provided by locating the safety-related instruments to maintain instrument channel separation within the spent fuel pool area, and to utilize inherent shielding from missiles provided by existing recesses and corners in the spent fuel pool structure.

1.2 Qualification

The level instrument channels shall be reliable at temperature, humidity, and radiation levels consistent with the spent fuel pool water at saturation conditions for an extended period.

1.3 Power supplies: Instrumentation channels shall provide for power connections from sources independent of the plant alternating current (ac) and direct current (dc) power distribution systems, such as portable generators or replaceable batteries. Power supply designs should provide for quick and accessible connection of sources independent of the plant ac and dc power distribution systems. Onsite generators used as an alternate power source and replaceable batteries used for instrument channel power shall have sufficient capacity to maintain the level indication function until KEPCO & KHNP 10

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 4-1 Post-Fukushima NRC Recommendations and Requirements (7 of 18)

NTTF NRC Recommendations/Requirements in SECY-Applicable DCD Rec. 11-0093, SECY-11-0137, SECY-12-0025, SECY APR1400 Design COL Action Note Section No 0095, EA-12-049, EA-12-051 offsite resource availability is reasonably assured.

1.4 Accuracy

The instrument shall maintain its designed accuracy following a power interruption or change in power source without recalibration.

1.5 Display

The display shall provide on-demand or continuous indication of spent fuel pool water level.

2. The spent fuel pool instrumentation shall be maintained available and reliable through appropriate development and implementation of a training program. Personnel shall be trained in the use and the provision of alternate power to the safety-related level instrument channels.

8 Strengthening and integration of emergency NA NA COL 19.3(7) operating procedures, severe accident management guidelines (SAMGs), and extensive damage mitigation guidelines (NTTF Recommendations)

1. Order licensees to modify the EOP technical guidelines (required by Supplement 1, Requirements for Emergency Response Capability,to NUREG-0737, issued January 1983 (GL 82-33), to (1) include EOPs, SAMGs, and EDMGs in an integrated manner, (2) specify clear command and control strategies for their implementation, and (3) stipulate appropriate qualification and training for those who make decisions during emergencies.

KEPCO & KHNP 11

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 4-1 Post-Fukushima NRC Recommendations and Requirements (8 of 18)

NTTF NRC Recommendations/Requirements in SECY-Applicable DCD Rec. 11-0093, SECY-11-0137, SECY-12-0025, SECY APR1400 Design COL Action Note Section No 0095, EA-12-049, EA-12-051

2. Modify Section 5.0, Administrative Controls,of the Standard Technical Specifications for each operating reactor design to reference the approved EOP technical guidelines for that plant design.

3. Order licensees to modify each plants technical specifications to conform to the above changes.

4. Initiate rulemaking to require more realistic, hands-on training and exercises on SAMGs and EDMGs for all staff expected to implement the strategies and those licensee staff expected to make decisions during emergencies, including emergency coordinators and emergency directors.

NA NA COL 19.3(8) 9.3 Emergency Preparedness (SECY-12-0025, DCD RAI 644-6516)

Communications

1. Provide an assessment of the current communications systems and equipment used during an emergency event to identify any enhancements that may be needed to ensure communications are maintained during a large scale natural event meeting the conditions described above. The assessment should:

Identify any planned or potential improvements to existing onsite communications systems and their required normal and/or backup power supplies, KEPCO & KHNP 12

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 4-1 Post-Fukushima NRC Recommendations and Requirements (9 of 18)

NTTF NRC Recommendations/Requirements in SECY-Applicable DCD Rec. 11-0093, SECY-11-0137, SECY-12-0025, SECY APR1400 Design COL Action Note Section No 0095, EA-12-049, EA-12-051

Identify any planned or potential improvements to existing offsite communications systems and their required normal and/or backup power supplies,

Provide a description of any new communications system(s) or technologies that will be deployed based upon the assumed conditions described above, and

Provide a description of how the new and/or improved systems and power supplies will be able to provide for communications during a loss of all ac power,

2. Describe any interim actions that have been taken or are planned to be taken to enhance existing communications systems power supplies until the communications assessment and the resulting actions are complete,

3. Provide an implementation schedule of the time needed to conduct and implement the results of the communications assessment.

NA NA COL 19.3(9) 9.3 Staffing

1. Provide an assessment of the onsite and augmented staff needed to respond to a large scale natural event meeting the conditions described above. This assessment should include a discussion of the onsite and augmented staff available to implement the strategies as discussed in the emergency plan and/or described in plant operating procedures.

KEPCO & KHNP 13

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 4-1 Post-Fukushima NRC Recommendations and Requirements (10 of 18)

NTTF NRC Recommendations/Requirements in SECY-Applicable DCD Rec. 11-0093, SECY-11-0137, SECY-12-0025, SECY APR1400 Design COL Action Note Section No 0095, EA-12-049, EA-12-051 The following functions are requested to be assessed:

How onsite staff will move back-up equipment (e.g., pumps, generators) from alternate onsite storage facilities to repair locations at each reactor as described in the order regarding the NTTF Recommendation 4.2. It is requested that consideration be given to the major functional areas of NUREG-0654, Table B¬1 such as plant operations and assessment of operational aspects, emergency direction and control, notification/ communication, radiological accident assessment, and support of operational accident assessment, as appropriate.

New staff or functions identified as a result of the assessment.

Collateral duties (personnel not being prevented from timely performance of their assigned functions).

2. Provide an implementation schedule of the time needed to conduct the onsite and augmented staffing assessment. If any modifications are determined to be appropriate, please include in the schedule the time to implement the changes.

3. Identify how the augmented staff would be notified given degraded communications capabilities.

KEPCO & KHNP 14

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 4-1 Post-Fukushima NRC Recommendations and Requirements (11 of 18)

NTTF NRC Recommendations/Requirements in SECY-Applicable DCD Rec. 11-0093, SECY-11-0137, SECY-12-0025, SECY APR1400 Design COL Action Note Section No 0095, EA-12-049, EA-12-051

4. Identify the methods of access (e.g., roadways, navigable bodies of water and dockage, airlift, etc.) to the site that are expected to be available after a widespread large scale natural event.

5. Identify any interim actions that have been taken or are planned prior to the completion of the staffing assessment.

6. Identify changes that have been made or will be made to your emergency plan regarding the on-shift or augmented staffing changes necessary to respond to a loss of all ac power, multi-unit event, including any new or revised agreements with offsite resource providers (e.g., staffing, equipment, transportation, etc.).

NA NA NA

- Filtration of Containment Vents The staff is considering requiring the filtration of containment vents to reduce the spread of radioactive contamination during a beyond-design-basis event.

The staff plans to provide the Commission a notation vote paper on these policy issues in July 2012.

At this time, the staff is proposing regulatory action to require that all operating BWR facilities with Mark I and Mark II containments have a reliable hardened venting capability, without filters, for events that can lead to core damage.

- Loss of Ultimate Heat Sink (SECY-12-0025)

1. Include UHS systems in the reevaluation and walkdowns of site-specific seismic and flooding hazards using the methodology described in KEPCO & KHNP 15

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 4-1 Post-Fukushima NRC Recommendations and Requirements (12 of 18)

NTTF NRC Recommendations/Requirements in SECY-Applicable DCD Rec. 11-0093, SECY-11-0137, SECY-12-0025, SECY APR1400 Design COL Action Note Section No 0095, EA-12-049, EA-12-051 SECY-11-0137, and identify actions that have NA NA COL 19.3(1) and been taken, or are planned, to address plant- 19.3(2) specific issues associated with the updated seismic and flooding hazards in conjunction with the resolution of NTTF Recommendations 2.1 and 2.3.

See NTTF 19.3.2.3 COL 19.3(3),

2. Incorporate the loss of UHS as a design Recommendation 19.3(4), 19.3(5),

assumption in the resolution of station blackout 4.2 in Subsection and 19.3(17) rulemaking activities in conjunction with the 5.1.2 resolution of NTTF Recommendation 4.1.

See NTTF 19.3.2.3 COL 19.3(3),

3. Provide mitigating measures for beyond-design-Recommendation 19.3(4), 19.3(5),

basis external events to also include a loss of 4.2 in Subsection and 19.3(17) access to the normal UHS in conjunction with 5.1.2 the resolution of NTTF Recommendation 4.2.

NA NA Refer to Tier 2

4. Include UHS systems in the reevaluation of site-Recommendation specific natural external hazards, and identify actions that have been taken, or are planned, to address plant-specific issues associated with the updated hazards in conjunction with the resolution of the new Tier 2 Recommendation 2.1 activity described in Enclosure 3, Other Natural External Hazards.

Tier 2 (Actions do not require long-term study and can be initiated when sufficient technical information and applicable resources become available) 2.1 Other External Events Protections No Action NA NA (SECY-12-0025)

1. Continue stakeholder interactions to discuss the technical basis and acceptance criteria for conducting a reevaluation of site-specific KEPCO & KHNP 16

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 4-1 Post-Fukushima NRC Recommendations and Requirements (13 of 18)

NTTF NRC Recommendations/Requirements in SECY-Applicable DCD Rec. 11-0093, SECY-11-0137, SECY-12-0025, SECY APR1400 Design COL Action Note Section No 0095, EA-12-049, EA-12-051 external natural hazards. These interactions will also help to define guidelines for the application of current regulatory guidance and methodologies being used for early site permit and combined license reviews to the reevaluation of hazards at operating reactors.

2. Develop and issue a request for information to licensees pursuant to 10 CFR 50.54(f) to (1) reevaluate site-specific external natural hazards using the methodology discussed in Item 1 above, and (2) identify actions that have been taken, or are planned, to address plant-specific issues associated with the updated natural external hazards (including potential changes to the licensing or design basis of a plant).

3. Evaluate licensee responses and take appropriate regulatory action to resolve issues associated with updated site-specific natural external hazards.

7 SFP Makeup Capability (NTTF 7.2, 7.3, 7.4, and No Action NA NA 7.5)

(NTTF Recommendations) 7.2 Order licensees to provide safety-related ac electrical power for the spent fuel pool makeup system.

7.3 Order licensees to revise their technical No Action NA NA specifications to address requirements to have one train of onsite emergency electrical power operable for spent fuel pool makeup and spent fuel pool instrumentation when there is irradiated fuel in the spent fuel pool, regardless of the operational mode of the reactor.

KEPCO & KHNP 17

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 4-1 Post-Fukushima NRC Recommendations and Requirements (14 of 18)

NTTF NRC Recommendations/Requirements in SECY-Applicable DCD Rec. 11-0093, SECY-11-0137, SECY-12-0025, SECY APR1400 Design COL Action Note Section No 0095, EA-12-049, EA-12-051 7.4 Order licensees to have an installed No Action NA NA seismically qualified means to spray water into the spent fuel pools, including an easily accessible connection to supply the water (e.g.,

using a portable pump or pumper truck) at grade outside the building.

7.5 Initiate rulemaking or licensing activities or both No Action NA NA to require the actions related to the spent fuel pool described in detailed recommendations 7.1-7.4.

9.3 Emergency preparedness regulatory actions (the No Action NA NA remaining portions of Recommendation 9.3, with the exception of Emergency Response Data System (ERDS) capability addressed in Tier 3)

1. Engage stakeholders to inform the development of acceptance criteria for the licensee examination of planning standard elements related to the recommendations, and

2. Develop and issue an order to address those changes necessary in emergency plans to ensure adequate response to SBO and multiunit events specific to (1) adding guidance to the emergency plan that documents how to perform a multiunit dose assessment, (2) conduct periodic training and exercises for multiunit and prolonged SBO scenarios, (3) practice (simulate) the identification and acquisition of offsite resources, to the extent possible, and (4) ensure that EP equipment and facilities are sufficient for dealing with multiunit and prolonged SBO scenarios.

KEPCO & KHNP 18

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 4-1 Post-Fukushima NRC Recommendations and Requirements (15 of 18)

NTTF NRC Recommendations/Requirements in SECY-Applicable DCD Rec. 11-0093, SECY-11-0137, SECY-12-0025, SECY APR1400 Design COL Action Note Section No 0095, EA-12-049, EA-12-051 Tier 3 (Those NTTF Recommendations that require further staff study to support a regulatory action) 2.2 Ten-year confirmation of seismic and flooding No Action NA NA hazards (dependent on Recommendation 2.1)

Initiate rulemaking to require licensees to confirm seismic hazards and flooding hazards every 10 years and address any new and significant information. If necessary, update the design basis for SSCs important to safety to protect against the updated hazards.

3 Potential enhancements to the capability to No Action NA NA prevent or mitigate seismically-induced fires and floods (long-term evaluation)

The Task Force recommends, as part of the longer term review, that the NRC evaluate potential enhancements to the capability to prevent or mitigate seismically induced fires and floods.

5.2 Reliable hardened vents for other containment No Action NA NA designs (long-term evaluation)

Reevaluate the need for hardened vents for other containment designs, considering the insights from the Fukushima accident. Depending on the outcome of the reevaluation, appropriate regulatory action should be taken for any containment designs requiring hardened vents.

6 Hydrogen control and mitigation inside No Action NA NA containment or in other buildings (long-term evaluation)

The Task Force recommends, as part of the longer term review, that the NRC identify insights about hydrogen control and mitigation inside containment or in other buildings as additional information is revealed through further study of the Fukushima Dai-ichi accident.

KEPCO & KHNP 19

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 4-1 Post-Fukushima NRC Recommendations and Requirements (16 of 18)

NTTF NRC Recommendations/Requirements in SECY-Applicable DCD Rec. 11-0093, SECY-11-0137, SECY-12-0025, SECY APR1400 Design COL Action Note Section No 0095, EA-12-049, EA-12-051 9.1 & Emergency preparedness (EP) enhancements for No Action NA NA 9.2 prolonged SBO and multiunit events (dependent on availability of critical skill sets) 9.1 Initiate rulemaking to require EP enhancements for multiunit events in the following areas:

personnel and staffing

dose assessment capability

training and exercises

equipment and facilities 9.1 Initiate rulemaking to require EP enhancements for prolonged SBO in the following areas:

communications capability

ERDS capability

training and exercises

equipment and facilities 9.3 ERDS capability (related to long-term evaluation No Action NA NA Recommendation 10)

Order licensees to do the following until rulemaking is complete:

Maintain ERDS capability throughout the accident.

10 Additional EP topics for prolonged SBO and No Action NA NA multiunit events (long-term evaluation) 10.1 Analyze current protective equipment requirements for emergency responders and guidance based upon insights from the accident at Fukushima.

10.2 Evaluate the command and control structure and the qualifications of decision-makers to ensure that the proper level of authority and oversight KEPCO & KHNP 20

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 4-1 Post-Fukushima NRC Recommendations and Requirements (17 of 18)

NTTF NRC Recommendations/Requirements in SECY-Applicable DCD Rec. 11-0093, SECY-11-0137, SECY-12-0025, SECY APR1400 Design COL Action Note Section No 0095, EA-12-049, EA-12-051 exists in the correct facility for a long-term SBO or multiunit accident or both.

Concepts such as whether decision-making authority is in the correct location (i.e., at the facility), whether currently licensed operators need to be integral to the ERO outside of the control room (i.e., in the TSC), and whether licensee emergency directors should have a formal license qualification for severe accident management.

10.3 Evaluate ERDS to do the following:

Determine an alternate method (e.g., via satellite) to transmit ERDS data that does not rely on hardwired infrastructure that could be unavailable during a severe natural disaster.

Determine whether the data set currently being received from each site is sufficient for modern assessment needs.

Determine whether ERDS should be required to transmit continuously so that no operator action is needed during an emergency.

11 EP topics for decision-making, radiation No Action NA NA monitoring, and public education (long-term evaluation) 11.1 Study whether enhanced onsite emergency response resources are necessary to support the effective implementation of the licensees KEPCO & KHNP 21

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 4-1 Post-Fukushima NRC Recommendations and Requirements (18 of 18)

NTTF NRC Recommendations/Requirements in SECY-Applicable DCD Rec. 11-0093, SECY-11-0137, SECY-12-0025, SECY APR1400 Design COL Action Note Section No 0095, EA-12-049, EA-12-051 emergency plans, including the ability to deliver the equipment to the site under conditions involving significant natural events where degradation of offsite infrastructure or competing priorities for response resources could delay or prevent the arrival of offsite aid.

11.2 Work with FEMA, States, and other external stakeholders to evaluate insights from the implementation of EP at Fukushima to identify potential enhancements to the U.S. decision-making framework, including the concepts of recovery and reentry.

11.3 Study the efficacy of real-time radiation monitoring onsite and within the EPZs (including consideration of ac independence and real-time availability on the Internet).

11.4 Conduct training, in coordination with the appropriate Federal partners, on radiation, radiation safety, and the appropriate use of KI in the local community around each nuclear power plant.

12.1 Reactor Oversight Process modifications to No Action NA NA reflect the recommended defense-in-depth framework (dependent on Recommendation 1)

Expand the scope of the annual reactor oversight process (ROP) self-assessment and biennial ROP realignment to more fully include defense-in-depth considerations.

12.2 Staff Training on Severe Accidents and Resident No Action NA NA Inspector Training on SAMGs (dependent on Recommendation 8)

Enhance NRC staff training on severe accidents, including training resident inspectors on SAMGs.

KEPCO & KHNP 22

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 5.0 STRATEGIES TO ACTION ITEMS FROM FUKUSHIMA DAI-ICHI EVENTS 5.1 Tier 1 Items 5.1.1 Recommendation 2.1 - Seismic and Flooding Re-Evaluation Seismic and flooding re-evaluation is the responsibility of COL applicant. The COL applicant will confirm that the site-specific design criteria for seismic and flood are met. It is expected that the APR1400 will satisfy the seismic requirements since it is designed to meet Central and Eastern United States (CEUS) seismic requirements. Also, for dry sites, the APR1400 will not have a problem in regard to flooding.

However, for wet sites, flood protection may be necessary depending on the location of FLEX equipment.

Therefore, the COL applicant will also address the flood requirements for wet sites.

5.1.2 Recommendations 4.1 and 4.2 - SBO and FLEX 5.1.2.1 Introduction This subsection summarizes the APR1400 diverse and flexible coping (FLEX) strategies for the beyond-design-basis external event (BDBEE), extended loss of all ac power (ELAP) concurrent with loss of normal access to ultimate heat sink (LUHS). The purpose of establishing the FLEX strategies is to maintain core cooling, spent fuel pool (SFP) cooling, and to ensure containment functions and structural integrity through containment cooling.

The core cooling safety function includes maintaining core cooling, reactor coolant system (RCS) inventory, RCS boration, and key reactor instrumentation. The containment cooling safety function includes maintaining containment pressure control, and key containment instrumentation. The SFP cooling safety function includes maintaining SFP cooling and key SFP instrumentation.

NTTF Recommendation 4 (Reference 1) recommends that all operating and new reactor designs enhance SBO mitigation capability for BDBEEs. Recommendation 4.1 outlines minimum coping times for SBO events. Recommendation 4.2 recommends that licensees provide reasonable protection from BDBEEs and add any additional equipment necessary to address multiunit events. This report addresses both Recommendation 4.1 and 4.2 through the baseline coping strategies.

5.1.2.2 Baseline Coping Capability The guidance for developing, implementing, and maintaining mitigation strategies from JLD-ISG-2012-01 (Reference 7) and the methodology to establish baseline coping capability from NEI 12-06 (Reference 8) were considered in developing the APR1400 FLEX strategies and evaluating the resultant baseline coping capability after the BDBEE.

The APR1400 FLEX strategies follow a three-phase approach as required in the Order EA-12-049 (Reference 5). The three phases are:

Phase 1 - Initial response phase using installed equipment

Phase 2 - Transition phase using FLEX equipment and consumables

Phase 3 - Indefinite sustainment of these functions using offsite resources.

The APR1400 baseline coping capability to maintain core cooling, SFP cooling, and containment cooling after the BDBEE are addressed in the following sections, along with the FLEX strategies and supporting analyses.

KEPCO & KHNP 23

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 5.1.2.3 Operational Strategy for Core Cooling The APR1400 core cooling capability to cope with the BDBEE, ELAP concurrent with LUHS, is addressed for all of the following operation modes.

a. Full-power operation

b. Low-power operations and shutdown conditions with steam generators (SGs) available

c. Shutdown conditions with SGs not available 5.1.2.3.1 FLEX Strategy for Full-Power Operation The initiating event is assumed to be a loss of offsite power (LOOP) with concurrent loss of all ac power and LUHS during full-power operation. Based on the analysis performed, the APR1400 design includes consideration of the following event sequence to address FLEX strategy for full-power operation:

Phase 1: 0 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Phase 2: 8 to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Phase 3: Indefinite time period following Phase 2 The timeline of the APR1400 FLEX strategy for full-power operation is shown in Figure 5-1 and the detailed sequence of events is tabulated in Table 5-1. The following are the operational strategies for each phase.

5.1.2.3.1.1 Phase 1: Coping with Installed Plant Equipment (0 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ) 5.1.2.3.1.1.1 Phase 1-a: 0 to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months The main control room (MCR) operators may require up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to assess plant conditions, equipment and system availability, and to identify the event as an ELAP concurrent with LUHS.

During Phase 1-a, only the installed plant equipment is used for coping. Specifically, two turbine-driven auxiliary feedwater pumps (TDAFWPs) automatically start on an auxiliary feedwater actuation system (AFAS) signal to provide core cooling through the SGs. Auxiliary feedwater storage tanks (AFWSTs) are used to supply water to the TDAFWPs, and steam generated in the SGs is released through the main steam safety valves (MSSVs). Class 1E batteries supply dc power to essential instrumentation and control (I&C) equipment, and for the operation of the TDAFWPs. The RCS is maintained at hot standby condition by the natural circulation cooling (NCC) operation without any operator action during this phase.

The reactor coolant pump (RCP) seal integrity may be challenged, because both the seal injection water supply and component cooling water supply to the RCP thermal barrier heat exchanger are lost due to the event of ELAP concurrent with LUHS. The RCP seal leakage is assumed to be 94.64 L/min (25 gpm) per RCP, based on the evaluation report on the APR1400 KSB RCP seals (Reference 9).

5.1.2.3.1.1.2 Phase 1-b (1 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months )

Throughout Phase 1, the RCS can be maintained at hot standby condition by the NCC operation without any operator action. The TDAFWPs are powered from the Class 1E batteries in Trains C/D for 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months without load shedding, and up to 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months with load shedding started at 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The AFWSTs supply KEPCO & KHNP 24

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 water to the TDAFWPs for core cooling and steam generated in the SGs is released through the MSSVs.

3 The total volume of water required from the AFWSTs during this phase is approximately 529.96 m (140,000 gal), which is much lower than the capacity of AFWSTs (See Table 5-2). The passive components, MSSVs, are operable without electrical power. Based on the analysis, the water level in the core is found to be well above the top of the active core during this phase, even though the RCP seal leakage is assumed to occur from the beginning of the event. During this phase, additional cooling in MCR, electrical and I&C equipment rooms, and the TDAFWP rooms is found not to be required based on heatup calculations. The maximum temperature of each room is guaranteed not to exceed the allowable values in table below to assure operability of equipment. As a result, Phase 1 coping time for this event can be extended to more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months . The COL applicant is to assure the survivability of the equipment Maximum Allowable Room Temperature MCR 110°F 1)

TDAFWP Rooms 150°F Electrical and I&C equipment rooms 120°F

1) The operability of equipment and instruments is guaranteed until 16hours after loss of HVAC.

In the meantime, the operator prepares for the next phase as soon as diagnosis of the event is complete:

operator tries to connect the 480 V mobile gas turbine generator (GTG) to the 480 V Class 1E power system Train A or B, and a primary high-head FLEX pump to the safety injection system (SIS) for RCS inventory makeup. Two secondary FLEX pumps are also connected to the SG auxiliary feedwater (AFW) supply lines, one for each AFW line. The operator action is assumed to be finished within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the event. Because the Phase 1 coping time can be extended to more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months , the operator has sufficient additional time margin to prepare for Phase 2, even though the period of Phase 1 is assumed to be 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months in the core cooling strategy.

5.1.2.3.1.2 Phase 2: Coping with Installed Plant Equipment and Onsite Portable Resources (8 to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months )

As soon as preparation for Phase 2 is finished, the operator starts to cool down the RCS to a safe shutdown state, i.e., hot shutdown or cold shutdown (see Figure 5-1), using the installed plant equipment and/or the onsite FLEX equipment.

During Phase 2, two types of core cooling strategies can be applied:

a. Basic operational strategy using installed plant equipment such as TDAFWP and the auxiliary charging pump (ACP), and FLEX equipment such as 480 V mobile GTG

b. Contingency plan using only the onsite FLEX equipment, which is applied if the installed plant equipment is not operable Each of these strategies is described below.

5.1.2.3.1.2.1 Basic Operational Strategy In the basic operational strategy, the RCS is cooled down to and maintained at the hot shutdown (176.67 °C [350 °F]) using both of installed plant equipment such as TDAFWP and ACP, and the FLEX equipment such as 480 V mobile GTG.

The RCS is cooled down to the hot shutdown condition by feed-and-bleed operation through the secondary side of SG using the TDAFWPs and main steam atmospheric dump valves (MSADVs). The KEPCO & KHNP 25

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 AFWSTs with backup water source from the raw water tank (RWT) continue to supply water to the SGs using the TDAFWP, while each SG level is maintained between 25 to 40 percent wide range by on-off control of the auxiliary feedwater isolation valves.

Two 480 V, 1,000 kW, mobile GTGs are provided to meet N+1 requirement. One of the 480 V mobile GTGs is connected to the 480 V Class 1E power system Train A or B, and supplies power to the 125 Vdc battery charger, the 480 V load center, and the motor control center (MCC). With this Class 1E power, ACP, MSADVs, and essential I&C equipment are available during this phase. During this phase, additional cooling in MCR, electrical and I&C equipment rooms, and ACP room is not required based on heatup calculations. The maximum temperature of each room is guaranteed not to exceed the allowable values in table below to assure operability of equipment. The COL applicant is to assure the survivability of the equipment Maximum Allowable Room Temperature MCR 110°F ACP Rooms 150°F Electrical and I&C equipment rooms 120°F ACP is used to provide makeup water for maintaining RCS inventory and provide RCP seal cooling. The suction source for ACP is the boric acid storage tank (BAST) and in-containment refueling water storage tank (IRWST). The water volume required for RCS inventory makeup during Phase 2 is approximately 3

643.52 m (170,000 gal).

5.1.2.3.1.2.2 Contingency Plan In the contingency plan strategy, installed plant equipment is assumed to be inoperable even after connection of 480 V mobile GTG. In this case, the RCS is further cooled down to around 110 °C (230 °F) with SGs fed by the two secondary side FLEX pumps instead of the plant installed TDAFWPs. RCS makeup is carried out by a primary side high-head FLEX pump instead of an ACP.

If the installed plant equipment, ACP, is inoperable, RCS inventory makeup can be provided by a primary side high-head FLEX pump 189.27 L/min (50 gpm) positive displacement pump with operating pressure 2

of 105.46 kg/cm A (1,500 psia).

Two secondary FLEX pumps are also connected to the SG auxiliary feedwater (AFW) supply lines: one for each AFW line. The secondary FLEX pumps can be used to supply feedwater to SGs, when 2

TDAFWPs are unavailable. If the SG pressure is under 6.33 kg/cm A (90 psia) during this phase, the TDAFWPs are inoperable. In this case, RCS is further cooled down to around 110 °C (230 °F) by 2

depressurizing the SG to 1.03 kg/cm A (14.7 psia) using the MSADVs. During this time, the SG inventory 2

is provided by the secondary side FLEX pumps (total dynamic head [TDH] of 17.01 kg/cm A [242 psia] at the rated flow rate of 1,173.48 L/min [310 gpm]) with suction from AFWST and RWT. The N+1 requirement for FLEX equipment is met by deploying two primary high-head FLEX pumps and three secondary FLEX pumps on site.

Additionally, as RCS cooldown continues and RCS pressure decreases to the designed setpoint of safety injection tank (SIT) injection, the SITs automatically discharge 4,000 ppm borated water into the RCS for boration and inventory makeup.

KEPCO & KHNP 26

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 5.1.2.3.1.2.3 Common Strategy to Both the Basic Strategy and Contingency Plan As the TDAFWPs or the secondary FLEX pumps continue to feed the SG, the AFWST inventory is depleted. Then, the suction of the TDAFWPs is realigned to the RWTs. The fuel for the mobile GTG is supplied by gravity flow from the emergency diesel generator (EDG) fuel oil day tank(s). Once the mobile GTG is running, the existing diesel fuel oil transfer pump is used to make up day tanks from the underground EDG fuel oil storage tanks, each having a capacity of 7 days of EDG operation at its continuous rating. Connections are also provided to supply fuel from the EDG fuel oil day tank to the primary and secondary side FLEX pumps for operation during Phase 2.

Table 5-2 shows the water volumes available from the AFWST and RWT during Phase 2. Although the water source from the RWT should be shared with SFP cooling, the AFWST and RWT are evaluated to be sufficient for continuous NCC operation for up to 11 days (see Table B-3 in Appendix B).

After the plant is brought to the safe shutdown state, i.e., hot shutdown or cold shutdown, the 4.16 kV mobile GTG and other resources, such as cooling water and fuel, will also be prepared by the end of Phase 2. All of the operator actions to prepare Phase 3 will be finished by 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following the event.

Even though Phase 2 is assumed to last until 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in this core cooling strategy, the capacity of the onsite SG cooling, RCS makeup water, and GTG fuel sources show the duration of Phase 2 can be extended up to 11 days. Hence, the operator has sufficient time margin to prepare for Phase 3.

The specific storage location, mobilization, and other details for the FLEX pumps and mobile GTGs are COL items.

5.1.2.3.1.3 Phase 3: Coping with both Installed and Offsite Resources in Addition to the Onsite Equipment (after 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months )

In Phase 3, offsite resources including a 4.16 kV mobile GTG, fuel, and cooling water can be assumed to be available for long-term coping with the BDBEE. The 4.16 kV mobile GTG is used to restore Train A or B of 4.16 kV Class 1E power system. The plant is brought to cold shutdown, using the shutdown cooling system (SCS) if the ultimate heat sink (UHS) is available after 4.16 kV Class 1E power is restored. If not, the plant is maintained at the same safe shutdown state as in Phase 2.

In this case, the primary and secondary makeup water sources and fuel oil for the mobile GTGs are refilled from offsite resources. The details for the offsite resources will be provided by the COL applicant.

5.1.2.3.2 FLEX Strategy for Low-Power and Shutdown Operation with SGs Available 5.1.2.3.2.1 Strategy for Mode 1 through Mode 3 The NCC analysis result for the full-power FLEX strategy is still valid for operation in modes 1 through 3, i.e., lower-power operation, startup, and hot standby conditions, because it covers various states of the plant, including full-power operation through hot shutdown condition. Therefore, the same FLEX strategy as in the full-power operation can be also applied to the Mode 1 through Mode 3 operations.

5.1.2.3.2.2 Strategy for Mode 4 and Mode 5 with SGs Available In these operation modes, the SCS normally maintains the RCS between 176.67 °C (350 °F) (hot shutdown) and 54.44 °C (130 °F ) (cold shutdown), while the SGs are still available.

If the event an ELAP concurrent with LUHS occurs during these operation modes, the RCS is heated up KEPCO & KHNP 27

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 and pressurized due to the loss of the SCS.

If the RCS temperature is initially below the maximum RCS temperature requiring low-temperature overpressurization protection (LTOP), i.e., 136.11 °C (277 °F), the RCS pressure can be maintained 2

below the LTOP protection limiting pressure of 43.94 kg/cm A (625 psia) (20 percent of the RCS hydraulic 2

test pressure of 219.71 kg/cm A [3,125 psia]), because the LTOP relief valve installed in the SCS 2

automatically opens at the opening setpoint (38.51 kg/cm A [530 psig]). After the RCS temperature increases to the LTOP disable temperature (136.11 °C [277 °F]), the operator tries to isolate the RCS from the SCS by manually closing the SCS isolation valves. The operator action for the isolation of the SCS is finished before the RCS temperature exceeds the SCS entry temperature 176.67 °C (350 °F).

After that, the RCS overpressurization can be protected by pilot-operated safety relief valves (POSRVs).

After closing of the SCS isolation valves, the RCS temperature and pressure continue to increase, and eventually return to the hot standby condition. Then, the SG side feed-and-bleed operation can start cooling down the RCS, as described in the baseline cooling capability for ELAP and LUHS at full-power operation. Consequently, the full-power FLEX strategy can be also applied after the plant returns to hot standby condition.

5.1.2.3.3 FLEX Strategy for Shutdown Operation with SGs Not Available The APR1400 shutdown operations with SGs not available include the mode 5 reduced inventory operation and the mode 6 refueling operation. If the ELAP concurrent with LUHS occurs during the reduced inventory operation or refueling, decay heat can be removed from the core by the RCS feed-and-bleed operation.

The APR1400 FLEX strategy follows a three-phase approach as required in Order EA-12-049 (Reference 5).

The following sections are general descriptions of FLEX strategy for shutdown operation with SGs not available. Detailed procedures will be developed by the COL applicant. The APR1400 added a COL item in DCD Chapter 19.3 specifying that the COL applicant will develop shutdown risk processes and procedures, and verify the ability to deploy FLEX equipment to provide core cooling in shutdown operations with SGs not available.

5.1.2.3.3.1 Phase 1: Coping with Installed Plant Equipment During Phase 1, decay heat is removed by the latent heat resulting from water boiloff in the core. The APR1400 will take actions to proceduralize administrative controls to pre-stage FLEX equipment prior to entering a condition where the SGs cannot provide adequate cooling.

5.1.2.3.3.2 Phase 2: Coping with Installed Plant Equipment and Onsite Portable Resources In Phase 2, the plant can be maintained at cold shutdown by the RCS feed-and-bleed operation using the FLEX pump which is connected to the SIS injection line. The RCS inventory makeup is carried out by external injection using the primary side low-head FLEX pump with rated flow of 2,839.06 L/min (750 gpm),

which is sufficient capacity for removing decay heat and flushing the RCS. Prior to core uncovery, the primary side low-head FLEX pump must be aligned to take suction from the acceptable coolant source and deliver the coolant to the vessel. A mobile GTG is prepared to connect to Train A or Train B 480 V Class 1E ac power system to supply power to Class 1E battery.

Two primary low-head FLEX pumps are provided to meet the N+1 requirement.

The specific storage location, mobilization, and other details for the FLEX pumps and mobile GTGs are COL items.

KEPCO & KHNP 28

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 5.1.2.3.3.3 Phase 3: Coping with Both Installed Plant Equipment and Offsite Resources in Addition to Onsite Equipment In Phase 3, the 4.16 kV mobile GTG, fuel, and cooling water are available from offsite for long-term coping for the event. The 4.16 kV mobile GTG is used to restore Train A or Train B of 4.16 kV Class 1E power system. If the SCS is operable when the 4.16 kV Class 1E power is restored, the plant is cooled down to and maintained at cold shutdown by resuming the SCS operation. If not, the operator keeps the plant at the same safe shutdown state as in Phase 2, using the primary FLEX pump for RCS inventory makeup. The primary makeup water source and fuel oil for the mobile GTGs are refilled using offsite resources. Details for the offsite resources will be provided by the COL applicant.

5.1.2.3.4 Supporting Analysis for Core Cooling Supporting analyses have been performed using RELAP5/Mod 3.3 to confirm the APR1400 core cooling capability to cope with the BDBEE, ELAP concurrent with LUHS, according to the FLEX strategies.

5.1.2.3.4.1 Acceptance Criteria The following acceptance criteria based on the NEI 12-06, Section 3.2.1 (Reference 8) are applied to the supporting analysis for the operational strategy for core cooling during the BDBEE:

Core cooling is maintained.

No fuel failures occur.

The fulfillment of above criteria is determined by evaluating RCS key parameters, such as RCS and SG pressures, RCS temperature, and collapsed levels in the reactor vessel, core, and SG.

5.1.2.3.4.2 Analysis Conditions The following analysis conditions and assumptions are selected according to the requirements of NEI 12-06, Section 3.2.1.

The plant is assumed to operate at 100 percent rated power with no uncertainty for system parameters.

The initiating event is assumed to be ELAP concurrent with LUHS.

The reactor is assumed to be tripped automatically by the low RCP speed trip function of the RPS since the RCPs could not be provided with ac power.

The MSSVs are assumed to actuate automatically when the SG pressure exceeds the MSSV setpoints.

RCP seal leakage is assumed to be 94.64 L/min (25 gpm) per RCP. This causes loss of RCS inventory, which should be adequately compensated for preventing core uncovery.

The TDAFWPs are assumed to start automatically on receipt of an AFAS signal.

The decay heat conditions of ANSI/ANS-5.1-1979 are used for best-estimate simulation of the KEPCO & KHNP 29

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 FLEX strategy.

The operator is assumed to cool down the RCS by controlling MSADVs with a rate of 27.78 °C/hr (50 °F/hr) from 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the BDBEE.

The auxiliary charging pump (ACP) is assumed to supply borated water at the constant value of 166.56 L/min (44 gpm) for RCS makeup after 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the event. If the ACP is unavailable for supplying water to RCS, a primary high-head FLEX pump is used for providing adequate water for maintaining RCS inventory.

Four SITs inject 4,000 ppm borated water into RCS when RCS pressure decreases below the setpoints as designed.

Normal feedwater flow to the SGs is assumed to stop at the initiation of the BDBEE. Auxiliary feedwater flow supplies water to the SGs and is controlled to maintain SG level within the control band of 25 to 40 percent.

5.1.2.3.4.3 Analysis Results and Conclusion The APR1400 core cooling capability under the BDBEE, ELAP concurrent with LUHS, is analyzed using the RELAP5/Mod 3.3 code, according to the full-power operational strategy, consisting of the following three phases as described in Subsection 5.1.2.3.1.

Phase 1: 0 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

Phase 2: 8 to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

Phase 3: Indefinite time period following Phase 2 For the full-power operation case, two types of core cooling strategies, which are basic operational strategy and contingency plan, are analyzed. In the basic strategy, the RCS is cooled down to hot shutdown using both installed plant equipment (such as MSADV, TDAFWP, and ACP), and FLEX equipment (such as the mobile GTG). The contingency plan is prepared in case the installed plant equipment is inoperable even after connection of mobile ac power. In this case, the RCS is cooled down to cold shutdown using the secondary side FLEX pump. RCS makeup is carried out by the primary side high-head FLEX pump.

Based on the two types of cooling strategies employed, it is concluded that the plant can be maintained at hot standby condition during Phase 1 (0 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the BDBEE), and cooled down to hot shutdown or cold shutdown state during Phase 2 (8 to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ), using onsite resources. The same safe shutdown state is also be maintained during Phase 3 (after 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ), by continuing NCC operation with the cooling water source (for FLEX pumps) and fuel oil (for mobile GTGs) supplied from offsite resources.

If SCS is successfully restored after the 4.16 kV GTG is connected, the plant can be brought to and maintained at cold shutdown using the SCS instead of SG cooling.

Appendix A provides further details of these analyses for the operational strategies - basic operational strategy and contingency plan - in the aspect of core cooling capability during this event.

Appendix A also shows the APR1400 coping capability against ELAP concurrent with LUHS during low-power operations and shutdown conditions. Based on the evaluation results for the operation modes 1 through 4 with SGs available, the aforementioned full-power operation strategy is still valid for this KEPCO & KHNP 30

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 condition.

In the operation modes 4 and 5 with SGs available, the SCS normally maintains the RCS between 176.67 °C (350 °F) (hot shutdown) and 54.44 °C (130 °F ) (cold shutdown), while the SGs are still available. If the ELAP concurrent with LUHS occurs during these operation modes, the RCS is heated up and pressurized for a period due to the loss of the SCS. If the RCS temperature is initially below the maximum RCS temperature requiring the LTOP, i.e., 136.11 °C (277 °F), the RCS pressure can be 2

maintained below the LTOP limiting pressure of 43.94 kg/cm A (625 psia), because the LTOP relief valve 2

installed in the SCS automatically opens at the opening setpoint (38.51 kg/cm A [530 psig]). Once the RCS temperature reaches the LTOP disable temperature (136.11 °C [277 °F]), the operator isolates the RCS from the SCS by manually closing the SCS isolation valves. The operator action for isolation of the SCS is finished before the RCS temperature exceeds the SCS entry temperature (176.67 °C [350 °F]).

After that, the RCS overpressurization can be protected by POSRVs. After closing the SCS isolation valves, the RCS temperature and pressure continue to increase, and eventually return to the hot standby condition. The full-power FLEX strategy can be also applied after the plant returns to hot standby.

The specific storage location, mobilization, and other details for the FLEX pumps and mobile GTGs are COL items.

5.1.2.4 SFP Cooling This subsection outlines the operational strategy to maintain the SFP water level at a safe condition throughout the BDBEE. The APR1400 SFP conditions are analyzed for a number of postulated scenarios for the ELAP event. The scenario with ELAP following a seismic event is found to be the most limiting case due to the higher SFP inventory loss.

5.1.2.4.1 Strategy for SFP Cooling Based on the supporting analyses (see Subsection 5.1.2.4.2) to determine the bulk SFP heatup time and boiloff rate, for a worst-case full core offload, these analyses concluded the following:

The operators have approximately 26.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to restore cooling and/or makeup to the SFP in order to keep the spent fuel covered. Therefore, boiling of the SFP can be credited as the Phase 1 event mitigation method.

To maintain at least 3.05 m (10 ft) of water inventory over the fuel assemblies, makeup water to the SFP is provided within 15.36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

For Phases 2 and 3 of event mitigation, an SFP makeup rate of 493.2 L/min (130.3 gpm) is needed to match the initial boiloff rate. The boiloff rate decreases over time as the spent fuel decay heat decreases.

Based on this information, an overview of the spent fuel cooling mitigation strategies is provided in Table 5-4. Specific details of the SFP mitigation strategies for each phase are provided in the following subsections.

5.1.2.4.1.1 Phase 1: SFP Cooling (0 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months )

In Phase 1, action is taken to open the rollup door to the fuel handling area truck bay on the El. 100-0 of the auxiliary building, prior to the onset of boiling, to establish a vent path from the area for steam generated from the SFP. Based on the analyses, SFP boiling is calculated to occur no sooner than 2.0 hours0 days 0 hours 0 weeks 0 months after the ELAP event occurs, considering the most limiting plant condition, i.e., mode 6 with full core KEPCO & KHNP 31

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 offload and failure of non-seismic Category I piping.

No makeup water to the SFP is required and the water level is monitored. During this phase, a FLEX pump with external makeup water connections to the RWT is established.

The vent path for the spent fuel area that is established in Phase 1 is maintained open in Phases 2 and 3.

5.1.2.4.1.2 Phase 2: SFP Cooling (8 to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months )

For Phase 2 event mitigation, makeup is required to the SFP. Based on the analyses presented in Appendix B, a minimum flow rate of 493.2 L/min (130.3 gpm) is required to match the worst-case SFP boiloff rate. This SFP makeup flow requirement is bounded, however, by the SFP makeup flow requirement needed to mitigate the effect of loss of large area (LOLA) per 10 CFR 50.54(hh) (2). The self-powered (diesel-driven) FLEX, 1,893 L/min (500 gpm) and 757 L/min (200 gpm), SFP makeup pump and spray pump relied on to mitigate LOLA are therefore credited to mitigate the BDBEE. These pumps are provided to meet the N+1 requirement for a single-unit site and will meet 10 CFR 50 Appendix A, General Design Criterion (GDC) 2.

SFP Makeup Water Source In all operating modes, the raw water tank (RWT) can be used as the water source for SFP makeup.

The RWT contains sufficient water inventory for SFP makeup required for mode 6 operation (limiting),

3 which is 1,676 m (442,883 gal) as shown in Appendix B, Table B-3.

Flexible hoses, FLEX pump(s), fuel for FLEX pump(s), and any other equipment required for this strategy are normally located away from the auxiliary building (i.e., greater than 91.44 m [100 yards]), so that mobilization of the equipment for SFP makeup capability can occur within the most limiting time of 15.03 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months for mode 6 operation (boiloff time to 3.05 m [10 ft] above fuel rack top water level).

Makeup Water FLEX Pump Staging and Pump Discharge Connections In Phase 2, the FLEX pump described above is staged outside the auxiliary building.

The FLEX pump discharge hose is routed to one of the two permanent SFP makeup connections located outside the east wall of the auxiliary building, as shown in Figure 6-2. One primary connection location is mounted on the wall of auxiliary building adjacent to, and just south of, the emergency diesel generator (EDG) building, and the other connection is mounted on the wall of auxiliary building adjacent to, and just north of, the emergency diesel generator building. The alternate connection is close to the SFP makeup connections. An SFP spray connection is close to each SFP makeup connection.

Standpipes to SFP Area The FLEX pump connections are each connected to an independent, seismically qualified standpipe that runs inside the auxiliary building from the pump staging area to a location above the SFP at El. 156 ft.

The two standpipes for the SFP makeup pump and the two standpipes for the SFP spray pump are located at diverse locations in the auxiliary building and extend from the ground elevation to the SFP elevation. The standpipes are suitably separated on the same side of the auxiliary building. The standpipes have connections at the bottom at ground elevation and the connections are external to the auxiliary building. Operators are able to connect flexible hoses to the standpipes, which are supplied by a FLEX pump. The standpipes for SFP makeup have hard-piped connections to the SFP edge to allow water makeup to the pool. The standpipes for SFP spray have hard-piped connections to the spray headers to allow spray into the pool. Each spray header is equipped with a number of spray nozzles to KEPCO & KHNP 32

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 direct flow into SFP area. An isolation valve and a check valve are provided on each standpipe.

The simplified arrangements of these makeup and spray provisions are shown in Figures 6-2 and 6-3.

SFP Makeup FLEX Pumps The diesel-driven FLEX pumps are provided to supply SFP makeup and SFP spray at a rate of at least 1,893 L/min (500 gpm) and 757 L/min (200 gpm) at discharge pressure heads of 32 m (105 ft) and 32.6 m (107 ft), respectively.

FLEX Equipment Storage The flexible hoses, FLEX pump, FLEX pump fuel supply, and any other FLEX equipment required for this strategy are stored away from the auxiliary building (i.e., greater than 91.44 m [100 yards] away) so that mobilization of the equipment for SFP makeup capability can occur within the 15.36-hour period identified above. The specific storage location, mobilization, and other details for the FLEX equipment are COL items.

5.1.2.4.1.3 Phase 3: SFP Cooling (after 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months )

The APR1400 continues operating with Phase 2 strategies to provide makeup to the SFP in Phase 3. In Phase 3, makeup to the RWT is provided from offsite water sources by the COL applicant.

5.1.2.4.2 Supporting Analyses for the Operational Strategy for SFP Cooling SFP decay heat removal capacity has been evaluated to confirm that SFP cooling can be continued during and after the occurrence of a BDBEE resulting in an ELAP and LUHS.

5.1.2.4.2.1 Evaluation Conditions NEI 12-06, Section 3.2.1.6 defines the following SFP conditions as general criteria and baseline assumptions for SFP conditions:

All boundaries of the SFP are intact, including the liner, gates, transfer canals, etc.

Although sloshing may occur during a seismic event, the initial loss of SFP inventory does not preclude access to the refueling deck around the pool.

SFP cooling system is lost.

SFP heat load assumes the maximum design basis heat load as defined in Appendix B.

Initial SFP water level is assumed at El. 149 ft 0 in (elevation of SFP cleanup suction nozzle).

SFP inventory makeup starts when the water level reaches Level 2 defined in NEI 12-02 (Reference 12).

Water inventories:

3 3 The water inventory above top of fuel rack: 816.3 m (28,826 ft )

3 3 Total SFP inventory: 816.3 m (28,826 ft )

KEPCO & KHNP 33

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 5.1.2.4.2.2 Evaluation Results From the detailed analysis presented in Appendix B, Table B-1, it can be seen that the worst-case SFP cooling load occurs in mode 6 with a full core offload.

5.1.2.4.2.3 Conclusions Based on the SFP time to boil and makeup analysis provided above and in Appendix B, considering a worst-case full core offload, the conclusions are as follows:

The operators have approximately 29.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to restore cooling and/or makeup to the SFP in order to keep the spent fuel covered. Therefore, boiling of the SFP can be credited as the Phase 1 event mitigation method.

To maintain at least 3.05 m (10 ft) of water inventory over the fuel assemblies, makeup to the SFP is provided within 15.36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

For Phase 2 and 3 event mitigation, an SFP makeup rate of 493.2 L/m (130.3 gpm) is needed to match the initial boiloff rate. The boiloff rate decreases over time as the spent fuel decay heat decreases.

5.1.2.5 Containment Function and Structural Integrity This subsection presents the operational strategy for containment cooling to ensure the containment function and structural integrity throughout the BDBEE. The APR1400 containment cooling capability to cope with the BDBEE is addressed in this subsection.

5.1.2.5.1 Strategy for Containment Cooling All of the containment heat removal systems are not credited to operate in the containment pressurization following mass and energy releases from the RCS during the BDBEE. However, even in that condition, the containment conditions need to be maintained within the containment design pressure and the environmental qualification (EQ) temperature that ensure functions of the safety related equipment within the containment as well as the containment structural integrity.

The emergency containment spray backup system (ECSBS) is considered as a means to cool the containment atmosphere, consequently decreasing the containment pressure and temperature below the design limits in such a case that the containment pressure increases following the BDBEE.

The ECSBS is manually actuated when the containment pressure reaches the design pressure. The RWT water is used as the water source for supplying to the ECSBS. The RWT has a sufficient capacity to provide water for the containment cooling including for the core and the SFP cooling.

5.1.2.5.2 Supporting Analyses for Containment Cooling The containment design incorporates a prestressed concrete containment with a steel liner to house the nuclear steam supply system. The containment and associated systems are designed to safely withstand environmental conditions that may be expected to occur during the life of the plant, including both short-term and long-term effects following a DBA and beyond DBA.

In the supporting analysis for the containment cooling during the BDBEE, the RCP seal LOCA at full-power operation is chosen as the representative case of power operation and shutdown conditions with SGs available to demonstrate maintaining the containment pressure and temperature below the design limits through the FLEX strategy that using the ECSBS.

KEPCO & KHNP 34

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 The APR1400 added a COL item in DCD Chapter 19.3 specifying that the COL applicant is to develop the shutdown risk processes and procedures, and verify the ability to deploy FLEX equipment to provide containment cooling in shutdown operation with SGs not available.

5.1.2.5.2.1 Acceptance Criteria The following acceptance criteria are applied to ensure the containment function and structural integrity throughout the BDBEE.

2

Containment pressure is maintained below the containment design pressure of 4.22 kg/cm (60 psig) at a BDBEE.

Containment temperature is maintained at a value less than the EQ temperature of 182 °C (360 °F).

5.1.2.5.2.2 Analytical Methods and Assumptions Containment pressure and temperature at a BDBEE are analyzed using the GOTHIC (Generation of Thermal-Hydraulic Information for Containments) computer code. The GOTHIC containment model, which was developed for the containment response calculation to a LOCA described in DCD Tier 2 Section 6.2.1, is used to demonstrate the containment capability to cope with the BDBEE. The GOTHIC containment model with analysis methodology are described in Technical Report "LOCA Mass and Energy Release Methodology" (Reference 17) in detail.

The major assumptions used in the containment analysis are listed below:

The ECSBS is actuated when the pressure reaches the containment design pressure in the containment pressurization following a BDBEE.

The discharge flow from the RCS leakage is instantly mixed into the containment atmosphere and reaches thermal equilibrium within the containment volume during the transient.

Variance in local temperature within the containment is not assumed since the accident scenario is characterized by slow, but continuous containment pressurization.

5.1.2.5.2.3 Analytical Results and Conclusion During the full-power operation, no major pipe break is postulated inside the containment, but RCP seal leakage is assumed to be at a leak rate of 94.64 L/min (25 gpm) per RCP, a total of 378.5 L/min (100 gpm) for four RCPs.

2 The containment pressure reaches the design pressure of 4.22 kg/cm (60 psig) after approximately 16 days of the accident, then rapidly decreases and maintains at a low pressure by continuous spray through the ECSBS. The ECSBS FLEX pump provides the flow rate of 2,839 L/min (750 gpm) at a discharge pressure head of 200 m (656 ft).

The containment temperature reaches the highest temperature of 148 °C (298 °F), but which is well below the temperature limit of 182 °C (360 °F) for the environmental qualification of the safety related equipment within the containment. Figure 5-3 and Figure 5-4 show the containment pressure and temperature transients to the RCP seal LOCA, respectively.

From the analysis, it is noted that the containment pressure and temperature are maintained well below the design limits, thereby ensuring the functions of safety-related equipment within containment as well as the containment structural integrity during the BDBEE.

KEPCO & KHNP 35

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 5.1.2.5.3 ECSBS Operation for Containment Cooling Connection Line The connection line to the FLEX pump runs inside the containment building through the auxiliary building from the Siamese connection to the ECSBS nozzles, which is designed to Seismic Category I standpipe.

The connection line is divided into two quality groups. The line from the Siamese connection to the ECSBS isolation valve (V1013) is designed with Quality Group D and the line from V1013 to ECSBS nozzles is designed with Quality Group B.

A simplified drawing that identifies the flow path to deliver water to the ECSBS is schematically shown in Figure 6-7.

Preparation for ECSBS operation If the ECSBS operation needs to be used, the following tasks are established.

a. Mobilize ECSBS FLEX pump and connect the pump suction to the RWT outlet for water supply

b. Connect pump discharge to the Siamese connection of the ECSBS standpipe

c. Open the RWT isolation valve in yard and enter general access area in the AB and open ECSBS isolation valve (CS-V1013) 5.1.2.6 Support Systems 5.1.2.6.1 Electrical Systems This subsection describes the electrical strategies to support the FLEX items described above for NTTF 4.1 and 4.2.

As stated earlier, the BDBEE causes the unit to lose all ac power. The initial condition is assumed to be a LOOP at a plant site resulting from a BDBEE that affects the offsite power system either throughout the grid or at the plant with no prospect for recovery of offsite power for an extended period. All installed sources of emergency onsite ac power and alternate ac power sources are assumed to be unavailable and not imminently recoverable.

However, the installed electrical distribution system, including inverters and battery chargers, remain available provided they are protected in a manner consistent with current station design.

5.1.2.6.1.1 AC Power The APR1400 has one 4.16 kVac, 5,000 kW and two 480 Vac, 1,000 kW mobile GTGs for the N+1 requirement, and those mobile GTGs are designed to meet the load requirements as stated in Table 5-5.

(See Appendix C for a detailed breakdown of electrical loadings.) The 480 V mobile GTG is credited to power the Class 1E 480 V load centers during Phase 2, while the 4.16 kV mobile GTG is credited to power the Class 1E switchgear during Phase 3.

The 4.16 kV mobile GTG is connected to the 4.16 kV switchgear Train A (or B), and the 480 V mobile GTG is connected to 480 V load center Train A (or B). The provisions to connect these GTGs are incorporated in the APR1400 design. The 4.16 kV GTG powers the 4.16 kV switchgear, 480 V load center and MCC, 480 Vac / 125 Vdc battery charger, 125 Vdc battery, 125 Vdc / 120 Vac inverter, and 120 Vac distribution panel in Train A (or B). The 480 V mobile GTG powers the 480 V load center and MCC, 480 Vac / 125 Vdc battery charger, 125 Vdc battery, 125 Vdc / 120 Vac inverter, and 120 Vac distribution panel in Train A (or B).

KEPCO & KHNP 36

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 During Phase 1, the APR1400 takes credit for Train C or D to which the TDAFWP is connected, while during Phases 2 and 3, the APR1400 takes credit for Train A or B. The ACP is designed to be powered from both Train A and Train B, and the MSADV is designed to be powered from either Train A or Train B.

Therefore, during Phases 2 and 3, when the mobile GTG is connected to either Train A or Train B, the APR1400 can be maintained in a safe condition. During Phase 3, the shutdown cooling pump and heat exchanger are used to recover the plant.

5.1.2.6.1.2 DC Power The APR1400 does not use mobile dc power supplies.

During an ELAP, Class 1E 125 Vdc power is required for operation of 4.16 kV switchgears, 125 Vdc loads, 480 Vac MOVs and AOVs that are backed up by 125 Vdc batteries, I&C panels and shutdown system instrumentation, and 120 Vac loads that are inverted from 125 Vdc batteries.

Both Train A and B batteries have a capacity of 2,800 Ah and can supply dc power up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months without load shedding and an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months with load shedding. Train C and D batteries have a capacity of 8,800 Ah and can supply dc power up to 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months without load shedding. The first 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (Phase 1) after the onset of BDBEE, the capacities of all the batteries are sufficient to provide dc power to all essential loads necessary to perform their safety duties.

During Phase 2, a 480 V mobile GTG is connected to either Train A or Train B of the Class 1E load center to supply power and recharge respective batteries to fully charged condition.

Battery Qualification The safety-related batteries that are extended for use longer than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , with reduced discharge rate through load shedding, are not required to be additionally qualified for FLEX profiles since the NRC endorsed the NEI White Paper with clarifications in September 2013 (References 10 and 11).

5.1.2.6.1.3 Emergency Lighting Emergency dc lighting in areas such as the MCR and technical support center (TSC) / operational support center (OSC) is provided from the Class 1E batteries during Phase 1, and emergency ac lighting is provided from the mobile GTG during Phases 2 and 3.

Access to manual valves requires lighting, and access to instrumentation monitoring or equipment operation also requires lighting. Under this adverse condition, the APR1400 is designed to provide portable lighting (e.g., flashlights or headlamps) as necessary to perform essential functions.

5.1.2.6.1.4 Communications Design features are incorporated into onsite plant communication system to enhance emergency preparedness for BDBEEs associated with simultaneous LUHS. These are described below.

The APR1400 communication subsystems provide an independent and diverse mode of communications.

A failure of one subsystem does not affect the capability to communicate using the other system.

Electric power is provided to the communications subsystems from the non-Class 1E uninterruptible power system (UPS) with 1-hour capacity in normal operation. The wireless communication system is supplied from the dedicated emergency UPS with 16-hour capacity.

However, normal communications may be lost or hampered during an ELAP. In this condition, portable communication devices are provided to support interaction between personnel in the plant and those KEPCO & KHNP 37

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 providing overall command and control. Communication gear (satellite phones and radios) are also provided for onsite and offsite communications. This system provides an alternate communication path for outside connections. The satellite telephone equipment includes a roof-mounted antenna and transceiver.

5.1.2.6.2 Water Supply System The primary source of water for the core cooling function is the AFWST for the first 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and the RWT can be used thereafter for up to 11 days, if required (Table 5-2).

For the SFP makeup and spray function, the RWT is the source of water.

The detailed design of the RWT related with site specific data is the responsibility of COL applicant. The COL applicant will confirm, satisfy, or fulfill the specific design functional requirements of raw water tank including the associated instrument, capacity, location, flow path to on-site, the valve pit connected to FLEX equipment, and any other design features as described in DCD Section 19.3 in support of BDBEE mitigation strategies. Also, the COL applicant will confirm and ensure that the RWT and flow path to the FLEX equipment (structures, piping, components, and connections) are designed to be robust with respect to applicable hazards (e.g., seismic events, floods, high winds, and associated missiles).

5.1.2.6.3 Fuel Oil Supply System EDG fuel oil day tank and the underground 7-day fuel oil storage tanks are used for running the diesel-driven FLEX pumps. During Phase 3, fuel oil is provided from an offsite source. Table 5-6 provides a summary of the fuel oil demands during the three phases of this event.

The existing onsite EDG fuel oil storage tanks and associated diesel fuel oil day tanks have a capacity of at least 32 days to sustain the demand for fuel oil during full-power mode operation 5.1.2.7 Summary of APR1400 Mitigation Capability for FLEX The APR1400 baseline capability is sufficient to support the safety functions of core cooling, containment function, and SFP cooling after BDBEE, with simultaneous loss of all ac power and LUHS. However, FLEX equipment stored onsite (or offsite) will be used to support the mitigation of a BDBEE resulting in an ELAP and LUHS. The APR1400 mitigation capability of the BDBEE is summarized in Table 5-7, which is based on the NEI 12-06 (Reference 8) FLEX capability matrix table. This table outlines baseline and FLEX capabilities of the APR1400 to maintain safety functions of core cooling, containment, and SFP cooling.

5.1.2.8 Conformance with NRC/NEI Recommendations Conformance with JLD-ISG-2012-01 (Reference 7) is addressed in Table 5-8.

Conformance with NEI 12-06 (Reference 8) is addressed in Table 5-9.

Conformance with NEI 12-06, Tables D-1, D-2, and D-3 (Reference 8) is addressed in Table 5-10.

5.1.3 Recommendation 7.1 - SFP Instrumentation 5.1.3.1 Introduction Recommendation 7.1 is a Tier 1 recommendation that resulted in the issuance of NRC Order EA-12-051 (Reference 6). The Order modified licenses to require a reliable indication of the water level in associated KEPCO & KHNP 38

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel: (1) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement makeup water addition should no longer be deferred.

The APR1400 SFP water level instrumentation is consistent with the guidelines in NRC EA-12-051 (Reference 6), NEI 12-02 Rev. 1 (Reference 12), and JLD-ISG-2012-03 Rev. 0 (Reference 13) as described in the following subsection.

5.1.3.2 Basic Strategy The strategy for addressing NTTF 7.1 SFP instrumentation is described below.

5.1.3.2.1 Identification of Spent Fuel Pool Water Levels The following are the key spent fuel pool water levels:

Level 1: Level adequate to support operation of the normal SFP cooling system Indicated water level on either the primary or backup instrument channel of greater than El. 144 ft 0in (based on ensuring the open end of the normal suction lines does not become uncovered) plus the accuracy of the SFP water level instrument channel.

Level 2: Level adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck Indicated level on either the primary or backup instrument channel of greater than 3.05 m (+/-0.305 m)

(10 ft [+/- 1 ft]) above the top of the fuel storage racks. The 3.05 m (10 ft) criterion is conservative with regard to dose, in that the APR1400 DCD Subsections 9.1.3.1 and 9.1.3.3.4 indicate that dose would remain at or below 0.025 mSv (2.5 mrem/hr) at the surface of the water. This monitoring level provides reasonable assurance that there is adequate water level to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck. The elevation associated with this level is greater than 139 ft 8 in plus the accuracy of the SFP water level instrument channel, which is determined at the COL stage.

Level 3: Level where fuel remains covered and actions to implement makeup water addition should no longer be deferred Indicated level is on either the primary or backup instrument channel of greater than 0.305 m (1 ft) above the top of the fuel storage racks. The elevation associated with this level is greater than 129 ft 8 in plus the accuracy of the SFP water level instrument channel, which is determined at the COL stage. This monitoring level provides reasonable assurance that there is adequate water level above the stored fuel in the rack.

5.1.3.2.2 Instruments The design of the instruments is consistent with the guidelines of NRC JLDISG-2012-03 Rev. 0 and NEI 12-02 Rev. 1. Specifically, the channels are designed as described below.

Primary (fixed) instrument channel (Channel A)

KEPCO & KHNP 39

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 The primary instrument channel provides level indication through the use of guided wave radar (GWR) technology using the principle of time domain reflectometry (TDR). The instrument provides a single continuous span from above Level 1 to within 0.305 m (1 ft) of the top of the spent fuel racks.

Backup instrument channel (Channel B)

The backup instrument channel is identical to the primary channel and is a permanent, fixed channel. The backup instrument channel provides level indication through the use of GWR technology using the principle of TDR. The instrument provides a single continuous span from above Level 1 to within 0.305 m (1 ft) of the top of the spent fuel racks.

The primary and backup instrument channels provide continuous level indication over a minimum range from the high SFP alarm El. 154 ft 2 in plus the accuracy of the SFP water level instrument channel to the top of the spent fuel racks at El. 129 ft 8 in minus the accuracy of the SFP water level instrument channel.

5.1.3.2.3 Reliability Conformance with the guidelines of NRC JLD-ISG-2012-03 Rev. 0 and NEI 12-02 Rev. 1 provides reasonable assurance of the reliability of the primary and backup instrument channels, as described in the subsections below. The GWR design was selected due to its reliability.

5.1.3.2.4 Instrument Channel Design Criteria Instrument channel design is consistent with the guidelines of NRC JLD-ISG2012-03 Rev. 0 and NEI 12-02 Rev. 1.

Instrument channels consist of a corrosion- and radiation-resistant metal probe submerged in the pool and connected to a corresponding display/processor by coaxial cable. The probe spans the length of the measured range of pool levels.

The probe is seismically mounted. It is designed to operate in borated and non-borated water over the entire expected range of pool conditions from normal temperatures to boiling temperatures. Cables and connections are designed for expected radiation levels and environments of greater than 100 °C (212 °F) and 100 percent humidity. Probes, cables, connectors, and mounting hardware in the area of the SFP are designed to function after the effects of seismically induced sloshing.

In the SFP area, cables shall be routed in seismically mounted rigid metal conduit. Outside the pool area, cables shall be routed in seismically mounted rigid metal conduit, trays, or raceways. Display/processors shall be mounted in promptly accessible areas outside of the SFP area as defined in Subsection 5.1.3.2.7 of this report.

Channels shall be physically separated by routing instrument cables in separate conduits, trays, or raceways, locating sensors on opposite sides of the pool near the corners, etc. Physical channel separation is maintained down through and including each channel display/processor where convergence may be allowed so that display/processors can be located close to each other or side by side.

Movement of the probe during a seismic event does not damage the pool liner and does not result in contact with spent fuel. Indication remains reliable after a seismic event.

Minor debris buildup on the probe does not impact performance.

5.1.3.2.5 Arrangement The channels/probes of level instruments are separated to reduce the potential for falling debris or KEPCO & KHNP 40

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 missiles affecting both channels of instrumentation. This placement, coupled with separate routing paths for cables and use of rigid conduit, provides reasonable protection against falling debris and structural damage.

Instrument power is derived from the Class 1E 480 V MCC. The MCC is located in the auxiliary building.

The MCC is expected to be in a mild environment after a BDBEE and can be easily accessed from the MCR; therefore, personnel can promptly obtain readings from the display. This building provides adequate protection against the effects of temperature, flood, humidity, radiation, seismic events, and missile hazards.

5.1.3.2.6 Mounting Both the primary and backup systems are installed as seismic Category I to meet the NRC JLD-ISG-2012-03 and NEI 12-02 guidance requirements.

Other hardware stored in the SFP is evaluated to provide reasonable assurance that it does not adversely interact with the SFP instrument probes during a seismic event.

5.1.3.2.7 Qualification Design criteria will provide reasonable assurance of instrument channel reliability during normal, event, and post-event conditions for no fewer than 7 days or until offsite resources can be deployed. The combination of analyses, operating experience, and/or manufacturer testing of channel components includes practices that are used to validate design criteria and considers the following:

Post-event conditions in the area of instrument channel components used for all instrument components

Effects of shock and vibration on all instrument channel components used during and following any applicable event for installed components

Seismic effects on instrument channel components used during and following a potential seismic event for installed components Components in the area of the SFP are designed for the temperature, humidity, and radiation levels expected during normal, event, and post-event conditions for no fewer than 7 days post-event or until offsite resources can be deployed by the mitigating strategies resulting from Order 12-049, Order Modifying Licenses With Regard to Requirements for Mitigation for Beyond-Design-Basis External Events." Examples of post-event conditions that are considered are:

Radiological conditions for a normal refueling quantity of freshly discharged (100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months ) fuel with SFP water level within 0.305 m (1 ft) of the top of the SFP racks (Level 3)

Temperature of 100 °C (212 °F) and 100 percent relative humidity environment

Boiling water and steam environment

The mitigating strategies developed in response to NEI 12-06, Diverse and Flexible Coping Strategies (FLEX)

Equipment located in the SFP is qualified to withstand a total accumulated dose of expected lifetime at normal conditions plus accident dose received at post-event conditions with SFP water level within 0.305 m (1 ft) of the top of the fuel rack seated in the spent fuel pool (Level 3).

KEPCO & KHNP 41

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 The metal probe and cable in the spent fuel pool area are robust components that are not adversely affected by expected radiation, temperature, or humidity. The areas selected for display/processor installation are considered mild environments in which personnel access is not prohibited by radiation, temperature, or humidity, and are readily accessible by operators during or after a BDBEE.

Components of the instrument channels are qualified for shock and vibration using one or more of the following methods:

Components are supplied by manufacturers that implement commercial quality programs (such as ISO 9001, Quality Management Systems - Requirements) with shock and vibration requirements included in the purchase specification at levels commensurate with portable hand-held devices or transportation applications.

Components have a history of operational reliability in environments with significant shock and vibration loading, such as portable hand-held device or transportation applications.

Components are inherently resistant to shock and vibration loadings, such as cables.

Demonstration of seismic adequacy is achieved using one or more of the following methods:

Demonstration of seismic motion consistent with that of existing design basis loads at the installed location.

Substantial history of operational reliability in environments with significant vibration, such as for portable hand-held devices or transportation applications. Such a vibration design envelope shall be inclusive of the effects of seismic motion imparted to the components proposed at the location of the proposed installation.

Adequacy of seismic design and installation is demonstrated based on the guidance in Sections 7, 8, 9, and 10 of Institute of Electrical and Electronics Engineers (IEEE) Standard 344-2004, IEEE Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations (Reference 14).

Demonstration that proposed devices are substantially similar in design to models that have been previously tested for seismic effects in excess of the plant design basis at the location where the instrument is to be installed (g-levels and frequency ranges).

Seismic qualification using seismic motion consistent with that of existing design basis loading at the installation location.

In those cases where the commercial quality program does not address the seismic levels and frequencies, additional analysis or testing will be provided.

5.1.3.2.8 Independence The primary instrument channel is independent of the backup instrument channel. Independence is obtained by physical separation of components between channels and the separate use of Class 1E MCC. The two permanently mounted instruments in the pool are physically separated as described in Subsections 5.1.3.2.4 and 5.1.3.2.5.

5.1.3.2.9 Power Supplies The power supplies for the instrument channels are as follows:

KEPCO & KHNP 42

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1

Each instrument channel is normally powered from a 120 Vac 60 Hz distribution panel of Class 1E 480 Vac MCC to support continuous monitoring of SFP level.

On loss of normal 120 Vac power from the Class 1E 480 V MCC, each channels internal UPS automatically transfers instrument power to a dedicated backup battery. If normal ac power is restored, the UPS automatically transfers instrument power back to the normal ac power.

The dedicated backup batteries are sized to be capable of supporting continuous monitoring of SFP level for a minimum of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of operation. This provides adequate time until the onsite or offsite mobile GTG can supply the power by mitigating strategies resulting from the ELAP event.

Instrument accuracy and performance are not affected by restoration of power or restarting the processor.

5.1.3.2.10 Accuracy Accuracy is consistent with the guidelines of NRC JLD-ISG-2012-03 Rev. 0 and NEI 12-02 Rev. 1.

Accuracy and indication features are as follows:

Accuracy: The absolute system accuracy is better than +/-7.62 cm (+/- 3 in). This accuracy is applicable for normal conditions and the temperature, humidity, chemistry, and radiation levels expected for BDBEE conditions.

Trending: The display trends and retains data when powered from either normal or backup power.

Restoration after loss of power: The system automatically swaps to available power (backup battery power or external dc source) when normal power is lost. Neither the source of power nor system restoration impact accuracy. Previously collected data are retained.

Diagnostics: The system performs and displays the results of real-time information related to the integrity of the cable, probe, and instrument channel.

The above features provide reasonable assurance that trained personnel can easily determine when SFP level falls below each regulatory level (Levels 1, 2, and 3) without conflicting or ambiguous indication.

5.1.3.2.11 Testing Testing and calibration are consistent with the guidelines of NRC JLD-ISG-2012-03 Rev. 0 and NEI 12-02 Rev. 1.

The display/processor performs automatic in-situ calibration and automatically monitors for cable, connector, and probe faults using TDR technology. Channel degradation due to age or corrosion is not expected but can be identified by monitoring trends.

The COL applicant should develop the station procedures and preventive maintenance tasks to perform required surveillance testing, calibration, backup battery maintenance, functional checks, and visual inspections of the probes.

5.1.3.2.12 Display The primary and backup level instruments are continuously displayed in the main control room and remote shutdown room.

KEPCO & KHNP 43

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 The displays are consistent with the guidelines of NRC JLD-ISG-2012-03 Rev. 0 and NEI 12-02 Rev. 1.

5.1.3.2.13 Instrument Channel Program Criteria The COL applicant should perform training and address the SFP level instrumentation maintenance procedure development in accordance with the guidelines of NRC JLD-ISG-2012-03 Rev. 0 and NEI 12-02 Rev. 1 as described below.

Training The systematic approach to training (SAT) is used to identify the population to be trained and to determine both the initial and continuing elements of the required training. Training is completed prior to placing the instrumentation in service.

Procedures Procedures for maintenance and testing are developed using regulatory guidelines and vendor instructions.

BDBEE operation guidance also addresses the following:

A strategy to ensure SFP water addition is initiated at an appropriate time consistent with implementation of NEI 12-06 Rev. 1.

Restoration of non-functioning SFP level channels after an event. Restoration timing is consistent with the emergency condition. After an event, commercially available components that may not meet all qualifications may be used to replace components to restore functionality.

5.1.3.3 Conformance with Regulatory Recommendations Conformance with NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, Rev. 1 (Reference 12), which the NRC has endorsed in JLD-ISG-2012-03, Rev. 0 (Reference 13), is summarized in Table 5-11.

SFP decay heat removal capacity is evaluated complying with NEI 12-06, Section 3.2.1.6 and summarized in Subsection 5.1.2.4.2.

5.1.4 Recommendation 8 - Emergency Response For the APR1400, NRC Rulemaking is applied. The final rule and draft guidance is due to the Commission by March 11, 2016.

COL applicants will develop emergency operating procedures (EOPs), severe accident management guidelines (SAMGs), and extensive damage mitigation guidelines (EDMGs) for their units that comply with the NRC rule for onsite emergency response that was issued in the NRC Advance Notice of Proposed rd Rulemaking (ANPR) on April 18, 2012 and is planned to be finally issued by the 3 quarter of 2016 as described in SECY-12-0025 (Reference 3).

The APR1400 added a COL item in DCD Chapter 19.3 specifying that COL applicant address strengthening and integration of EOPs, SAMGs, and EDMGs.

KEPCO & KHNP 44

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 5.1.5 Recommendation 9.3 - Emergency Plan (only staffing and communications equipment portion in Tier 1) 5.1.5.1 Communication Equipment Considering the Request for Information (RFI) on NTTF Recommendation 9.3, "Emergency Preparedness" depicted in Enclosure 5 to SECY-12-0025 (Reference 15) and NEI 12-01(Reference 16),

the following design features are incorporated into the onsite plant communication system to enhance emergency preparedness for a BDBEE associated with simultaneous loss of all ac power and LUHS, in addition to the existing design features of the station communication system:

1. Addition of power sources for the wireless communication systems

2. Introduction of a satellite telephone link Regarding offsite communications, a new COL item is added in the DCD Chapter 19.3 specifying that COL applicants are to address the offsite communication requirements specified in Enclosure 5 to SECY-12-0025.

5.1.5.2 Staffing COL applicants that construct an APR1400 are responsible to conduct staffing evaluations for the unit in response to the emergency planning staffing provisions of Recommendation 9.3. A new COL item is added in DCD Chapter 19.3 specifying that COL applicants are to address staffing evaluations for the unit, considering the requested functions described in Recommendation 9.3, Items 1 through 4 and 6 (Reference 1), including those related to NTTF Recommendation 4.2.

5.2 Tier 2 Items 5.2.1 Recommendation 7.2 - Safety-Related ac Electrical Power for the SFP Makeup System NTTF Recommendation 7.2 is a Tier 2 recommendation that requests safety-related ac power for the SFP makeup system.

The APR1400 design provides the AFWSTs as a seismic Category I makeup water source for SFP. The makeup water is delivered to the SFP via component cooling water (CCW) makeup pumps, which are powered by safety-related ac power.

5.2.2 Recommendation 7.3 - Plant Technical Specifications NTTF Recommendation 7.3 is a Tier 2 recommendation that requests that Plant Technical Specifications require one train of emergency onsite electrical power to be operable for SFP makeup/instrumentation when there is irradiated fuel in the SFP, regardless of plant operating mode.

The APR1400 Technical Specifications require at least one EDG and one CCW makeup pump to be operable in all modes. The safety-related CCW makeup pumps are powered by the EDGs. The CCW makeup is capable of providing makeup to the SFP.

Refer to the APR1400 DCD Tier 2, Chapter 16, Technical Specification 3.8.1.

5.2.3 Recommendation 7.4 - Seismically Qualified Spent Fuel Pool Spray System Recommendation 7.4 is a Tier 2 recommendation that requests that a seismically qualified means to spray water into SFPs be provided, including an easily accessible connection to supply water, such as KEPCO & KHNP 45

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 using a FLEX pump or pumper truck, at grade level outside of the building The APR1400 design provides a diverse spent fuel pool makeup and spray system as described in this report that meets the criteria specified in this recommendation.

5.2.4 Recommendation 7.5 - Spent Fuel Pool Actions Related to Recommendations 7.1 through 7.4 Considering Near-Term Report and Recommendations for Agency Actions Following the Events in Japan, dated July 12, 2011 (Reference 1), the SFP diverse makeup lines and spray lines of the APR1400 are designed to withstand a safe shutdown earthquake (SSE).

5.2.5 Recommendation 9.3 - Emergency Preparedness Regulatory Actions (remaining portions of Recommendation 9.3, except Emergency Response Data System - ERDS capability addressed in Tier 3)

COL applicants that construct an APR1400 will be implementing improved emergency preparedness activities, which the NTTF Recommendation intends.

In SECY-12-0095 (Reference 4), the NRC staff proposed a project plan for publication of an ANPR on emergency preparedness activities addresses the Tier 2 and Tier 3 components of NTTF Recommendation 9.3, as summarized in Table 4-1 of this Technical Report.

5.2.6 Recommendation 2.1 - Other External Events This was newly added as a Tier 2 issue in SECY-12-0025. As the intent of the recommendation is re-evaluation of the existing operating reactors against a set of new plant evaluation criteria and the other external events have been fully evaluated in the APR1400 DCD, neither KHNP nor COL applicants need take action on this recommendation.

5.3 Tier 3 Items (and Other Items) 5.3.1 Recommendation 2.2 - Ten-Year Confirmation of Seismic and Flooding Hazards (dependent on Recommendation 2.1)

This recommendation is only applicable to operating plants. Therefore, neither KHNP nor COL applicants will take any action on this recommendation.

5.3.2 Recommendation 3 - Potential Enhancements to the Capability to Prevent or Mitigate Seismically-Induced Fires and Floods (Long-Term Evaluation)

Neither KHNP nor COL applicants are required to take action because in SECY-12-0095 (Reference 4),

the NRC staff proposed that it will defer the evaluation of NTTF Recommendation 3 (Reference 1) until 2016.

5.3.3 Recommendation 5.2 - Reliable Hardened Vents for Other Containment Designs (Long-Term Evaluation)

The staff plans to defer consideration of venting for other containment designs (e.g., Mark III, ice condenser, and large dry containments) until the Commission reaches a decision on the need for severe accident venting and filtered venting for BWR Mark I and Mark II containments. Therefore, KHNP will not take action on this recommendation at this time.

KEPCO & KHNP 46

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 5.3.4 Recommendation 6 - Hydrogen Control and Mitigation inside Containment or in Other Buildings The NRC plans to develop and issue a rule on hydrogen control and mitigation inside the containment or in other buildings as a long-term issue. Currently, there is an insufficient information to substantiate changes to the current APR1400 design. Therefore, KHNP will not take action on this recommendation at this time.

5.3.5 Recommendations 9.1 and 9.2 - Emergency Preparedness (EP) Enhancements for Prolonged SBO and Multiunit Events The NRC plans to develop and issue a rule on emergency preparedness by 2016. No action is required at this time.

5.3.6 Recommendation 9.3 - ERDS Capability The NRC plans to develop and issue a rule on emergency preparedness by 2016. No action is taken.

5.3.7 Recommendation 10 - Additional EP Topic for Prolonged SBO and Multiunit Events The NRC plans to develop and issue a rule on emergency preparedness by 2016. No action is taken.

5.3.8 Recommendation 11 - EP Topics for Decision-Making, Radiation Monitoring, and Public Education The NRC plans to develop and issue a rule on emergency preparedness by 2016. No action is required at this time.

5.3.9 Recommendation 12.1 - Reactor Oversight Process Modifications to Reflect the Recommended Defense-in-Depth Framework The NRC staff proposed in SECY-12-0095 that the staff will defer action on Recommendation 12.1 until the Commission has provided staff guidance regarding Recommendation 1. No action is required at this time.

5.3.10 Recommendation 12.2 - Staff Training on Severe Accidents and Resident Inspector Training on SAMGs This recommendation is related to the NRC. No action is required at this time.

5.3.11 Additional Recommendations There are two additional recommendations as follows:

Emergency planning zone (EPZ)

Pre-staging of potassium iodide beyond 16,093 m (10 miles)

As these recommendations concern the fundamental issue of existing regulatory framework that should be developed by the NRC, no action is required.

KEPCO & KHNP 47

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-1 Sequence of Events for Core Cooling (Full-Power Operation) (1 of 2) TS KEPCO & KHNP 48

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 TS Table 5-1 Sequence of Events for Core Cooling (Full-Power Operation) (2 of 2)

KEPCO & KHNP 49

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-2 Water Volume Source and Requirements for SG Feedwater Phase 1 -

Tank Volume, Total 3 3 Total Volume Phase 1 & 2 - Total Volume Tank Quantity m /tank Volume, m 3 3 Required, m Required, m (gal)

(gal/tank) (gal)

(gal)

Auxiliary 2 1,819.61 3,639.22 529.96 2,849.69 Feedwater (480,690) (961,380) (140,000) (752,809)

Storage Primary source of water is Tank AFWST for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . RWT can (AFWST) be used after depletion of Raw Water 2 4,996.74 9,993.49 NA AFWSTs for up to 11 days Tank (RWT) (1,320,000) (2,640,000) following the event.

KEPCO & KHNP 50

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-3 FLEX Capability - Spent Fuel Pool Cooling Summary Safety Function Method Phase 1 Phase 2 and 3 Spent Fuel Pool Makeup through connection Analysis demonstrates that spent fuel Permanent connections for FLEX, self-powered Cooling to SFP makeup piping or heats up slowly and remains cooled by SFP makeup pump and SFP spray pump. (FLEX other suitable means (e.g., water inventory above the top of the connection locations and equipment are protected sprays) spent fuel rack. from the applicable hazards in NEI 12-06. They are designed to seismic Category I requirements. They either are located in seismic Category I structures or outside, above the ground).

Makeup with FLEX injection Analysis demonstrates that spent fuel Permanent connections to make up the SFP from source heats up slowly and remains cooled by RWT.

water inventory above the top of the spent fuel rack.

Vent pathway for steam Vent path from SFP area to environment Vent path established in Phase 1 is maintained established for removal of steam. (Rollup open to provide a vent path for steam.

door to the fuel handling area truck bay is opened prior to earliest predicted spent fuel pool time to boil.)

SFP Parameters SFP level Instruments powered by Class 1E MCC. On loss of normal 120 Vac power from the Class 1E 480 V MCC, each channels internal UPS The APR1400 design includes automatically transfers instrument power to a redundant, safety-related wide-range dedicated backup battery to support continuous level sensors in SFP that fulfill EA-12-051 monitoring of SFP level. If normal ac power is order. restored, the UPS will automatically transfer instrument power back to the normal ac power.

KEPCO & KHNP 51

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-4 480 V and 4.16 kV Mobile GTG Electrical Load Summary List (in kW)

Electrical Load Description Train A Train B Remark Loads 762.4 810.4 480V Mobile GTG Total (with 10% Rating: 1,000 kW 838.6 891.4 margin)

Loads 4,204.48 4,217.94 4.16 kV Mobile GTG Total (with 10% Rating: 5,000 kW 4,624.93 4,639.09 margin)

KEPCO & KHNP 52

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-5 Summary of Fuel Oil Demand (most limiting)

Fuel Oil Volume Fuel Oil Equipment Phase Purpose required Remark Source (Specification)

Liters (gal)

Phase 1 Core cooling NA EDG fuel oil GTG 29,072 7.57 L/min Power supply storage tank (480 V/1,000 (7,680) (2.0 gpm) and day tank kW)

Two secondary FLEX pumps (each 1,174 2,764 0.36 L/min Core cooling L/min, 160 m (730) (0.095 gpm)

[310 gpm, 525 ft])

One primary high-head FLEX Phase 2 pump (190 2,060 0.54 L/min (modes 1~4) RCS makeup 2 L/min, 17 kg/cm (544) (0.142 gpm)

A [50 gpm, 243 psia])

174

- One SFP (46) makeup FLEX SFP cooling pump (1,893 (No full core L/min [500 gpm]) An alternate offload)

- SFP spray (757 379 means of SFP L/min [200 gpm]) (100) makeup FLEX pump RCS makeup Phase 3 Resources external (COL)

SFP cooling KEPCO & KHNP 53

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-6 APR1400 FLEX Capability Summary (1 of 3)

Safety Function Method Capabilities FLEX Equipment Core Cooling Core Cooling (SGs Phase 1:

Use of installed equipment

Onsite self-powered and RCS available): Modes

NCC (TDAFWP-SG-MSSV-MSADV, ACP, primary high-head Inventory 1 through 5

TDAFWP-SG-MSSV- SIT, UHS, SCS) FLEX pump AFWST

Use of water supply (AFWST, RWT,

Onsite self-powered Phase 2: IRWST) secondary side FLEX

NCC

Use of a primary side high-head FLEX pump to directly supply

TDAFWP-AFWST-SG- pump if ACP is not available water to SG MSADV

Use of secondary side FLEX pumps if

ACP, SIT TDAFWPs are not available

480 V onsite mobile GT

FLEX pumps

Connection for FLEX pumps to supply G

Load shedding water

4.16 kV offsite mobile

480 V mobile GTGs

Use of UHS/SCS instead of the GTG

IRWST, AFWST NCC cooling with MSADVs and Phase 3: TDAFWPs after the 4.16 kV mobile

Same as Phase 2 GTG is connected, if UHS is restored

RWT

Offsite resources

4.16 kV mobile GTG Core Cooling (SGs Phase 1

Use of water supply (IRWST)

Onsite self-powered unavailable):

Decay heat is removed by

Vent steam through PZR manway primary side FLEX Modes 5 and 6 boiloff from the core

Use of FLEX pump pump to make up RCS

RCS makeup connections for a FLEX

480 V onsite mobile GT

Feed-and-bleed by external primary side pump G injection using FLEX pump Use of UHS/SCS instead of the NCC

4.16 kV offsite mobile

Available borated water cooling with MSADVs and TDAFWPs GTG source (COL 19.3(17)) after the 4.16 kV mobile GTG is

480 V mobile GTGs connected, if UHS is restored Phase 3

4.16kV mobile GTG

UHS/SCS

Available borated water source (COL 19.3(17))

Offsite resources KEPCO & KHNP 54

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-6 APR1400 FLEX Capability Summary (2 of 3)

Safety Function Method Capabilities FLEX Equipment RCS Inventory/

Low-leakage RCP seals

Use of ACP or primary side high-head

Onsite self-powered Boration (SGs (leakage assumed to be FLEX pump for RCS makeup with primary side FLEX available): Modes 94.64 L/m [25 gpm] per borated water

480 V onsite mobile 1 through 5 RCP)

SIT for boration GTG

Provide borated RCS

4.16 kV offsite mobile makeup GTG RCS Inventory/

RCS feed and bleed

Use of primary side low-head FLEX

Onsite self-powered Boration (SGs not pump for RCS makeup primary side FLEX available): Modes pump 5 and 6 Key

SG water level and

Instruments powered by Class 1E dc

Onsite mobile GTG Instrumentation pressure bus

AFWST water level

RCS hot leg (HL) and cold leg (CL) temperature

Pressurizer (PZR) water level and pressure

Core exit temperature Containment Containment

Containment Structure

Source of water (AFWST, RWT)

Onsite self-powered Integrity Cooling

ECSBS

ECSBS spray FLEX pumps to provide containment cooling, with hoses and couplings Key Containment

Containment pressure

Key instruments powered by Class 1E

None Instrumentation dc bus SFP Cooling SFP Cooling

SFP makeup and SFP

Use of installed equipment (RWT)

Onsite self-powered spray

Use of FLEX pumps FLEX pumps to makeup

SFP makeup lines and SFP spray lines SFP, with hoses and from yard connections to SFP couplings

Vent pathway for steam and condensed vapors from SFP area SFP Instruments

SFP level instrumentation

Two sets of wide range (Level 1 to 3)

Onsite mobile GTG safety-related, continuous SFP level instruments KEPCO & KHNP 55

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-6 APR1400 FLEX Capability Summary (3 of 3)

Safety Function Method Capabilities FLEX Equipment Support ac Power

Mobile ac power source

ac distribution system including inverters

One 4.16 kV mobile GTG /

Function

and battery chargers installed two 480 V mobile GTGs (N+1 requirement) dc Power

dc power source via battery

dc distribution system including inverters

One 4.16 kV mobile GTG /

chargers and battery chargers installed two 480 V mobile GTGs

dc distribution system (N+1 requirement)

HVAC

HVAC system for MCR ,

HVAC system for MCR, electrical and I&C

No cooling is necessary electrical and I&C equipment equipment rooms, ACP room, and TDAFW for MCR, electrical and rooms, ACP room, and pump room I&C equipment room, ACP TDAFWP room room and the TDAFWP room as the heatup temperature does not exceed the design temperature of the room during BDBEE Lighting

Emergency lighting systems

Emergency lighting powered from

Onsite mobile GTG to Class 1E 125 Vdc batteries recharge the Class 1E

Emergency portable lighting batteries Communication

Communication systems

Plant communication systems powered by

Onsite mobile GTG to Non-Class 1E UPS for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recharge the batteries

Emergency wireless communication supplying to Class 1E devices powered by dedicated emergency MCCs UPS for 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months Fuel Oil

Fuel oil system

Connection for onsite emergency diesel

Onsite EDG fuel oil generator fuel oil system storage tanks

Onsite EDG fuel oil day tanks

External resources Makeup Water

Makeup water source

AFWST for TDAFW pump and backup from

See the FLEX equipment RWT for core cooling, SFP

RWT for SFP makeup and SFP spray cooling for components operation

External water source KEPCO & KHNP 56

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-7 Conformance with JLD-ISG-2012-01, Rev. 0 (1 of 5)

JLD-ISG-2012-01 Rev. 0 APR1400 Section Summary 1.0 Evaluation of NEI 12-06, Section 4 describes the overall methodology for COL applicants are responsible to assess the site-External evaluating the impact of the hazards, described in Sections 5.0 specific external hazards in accordance with the Hazards through 9.0, on the deployment of the strategies to meet the guidance.

baseline coping capability.

2.0 Phased Order EA-12-049 requires a three-phase approach to mitigating The APR1400 FLEX strategy complies with the Approach beyond-design-basis events, with an initial response phase using guidance.

installed equipment, a transition phase using portable equipment and The APR1400 FLEX strategy to provide core and SFP consumables to provide core and spent fuel pool (SFP) cooling and cooling, and to maintain containment integrity when maintain the containment functions, and a third phase of indefinite ELAP and LUHS are assumed to occur sustainment of these functions using offsite resources. simultaneously, follows the three-phase approach as Maintenance of core and SFP cooling and containment functions requested in Order EA-12-049. The three-phase requires overlap between the initiating times for the phases with the operations consist of an initial response phase using duration for which each licensee can perform the prior phases. The installed equipment, a transition phase using FLEX NRC staff recognizes that for certain beyond-design-basis external equipment and consumables to provide core cooling, events, the damage state could prevent maintenance of key safety and a third phase of indefinite sustainment of these functions using the equipment intended for particular phases. Under functions using offsite resources.

such circumstances, prompt initiation of the follow-on phases to restore core and SFP cooling and containment functions is appropriate. If fuel damage occurs, the Severe Accident Management Guidelines should be used as guidance.

2.1 Initial The initial response phase will be accomplished using installed The APR1400 FLEX strategy complies with the Response equipment. guidance.

Phase Licensees should establish and maintain current estimates of their FLEX strategy for power operation and shutdown capabilities to maintain core and SFP cooling and containment mode with SGs available:

functions assuming a loss of alternate current (ac) electric power to Initial phase:

the essential and nonessential switchgear buses except for those fed During the initial response phase, it is assumed that by station batteries through inverters. all ac power and normal access to UHS are lost, but This estimate provides the time period in which the licensee should be the dc battery is available. Train C and D Class 1E able to initiate the transition phase and maintain or restore the key battery supplies dc power to essential I&C equipment, safety functions using portable onsite equipment. and TDAFWPs continue to feed SGs at least for 8 This estimate should be considered in selecting the storage locations hours following the event. Also, the steam generated for that equipment and the prioritization of resources to initiate their from SG is released though the passive safety valves, use. MSSVs. Therefore, NCC operation to maintain RCS at hot standby is possible without any operator action during this phase.

KEPCO & KHNP 57

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-7 Conformance with JLD-ISG-2012-01, Rev. 0 (2 of 5)

JLD-ISG-2012-01 Rev. 0 APR1400 Section Summary FLEX strategy for shutdown mode with SGs not available:

Initial phase:

Decay heat is removed by RCS inventory boiloff from core.

2.2 Transition The transition phase will be accomplished using portable equipment The APR1400 FLEX strategy complies with the Phase stored onsite. guidance.

The strategies for this phase must be capable of maintaining core FLEX strategy for power operation and shutdown cooling, containment, and spent fuel pool cooling capabilities mode with SGs available:

(following their restoration, if applicable) from the time they are Transition phase:

implemented until they can be supplemented by offsite resources in During this phase, RCS is cooled down to around the final phase. 176.67 °C (350 °F) using the installed plant The duration of the transition phase should provide sufficient overlap equipment, such as TDAFWP, MSADV, ACP, SIT, with both the initial and final phases to account for the time it takes to and/or FLEX equipment, such as 480 V mobile GTG install equipment and for uncertainties. and primary FLEX pump. If installed plant equipment is inoperable even after connection of mobile ac power, RCS is further cooled down to around 98.89 °C (210 °F) using secondary side FLEX pump. RCS makeup is carried by the primary side FLEX pump.

AFWST and RWT are consecutively used as onsite water sources to feed SGs. The transition phase can be extended to approximately 11 days.

Therefore, the duration of the transition phase provides sufficient overlap with final phase.

The initial phase overlaps for at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months with the transition phase, since dc battery is available until 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months following the event without load shedding.

FLEX strategy for shutdown mode with SGs not available:

Transition phase:

The plant is maintained at cold shutdown by the RCS feed-and-bleed operation using the primary side low-head FLEX pump KEPCO & KHNP 58

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-7 Conformance with JLD-ISG-2012-01, Rev. 0 (3 of 5)

JLD-ISG-2012-01 Rev. 0 APR1400 Section Summary 2.3 Final Phase The final phase will be accomplished using the portable equipment The APR1400 FLEX strategy complies with the stored onsite augmented with additional equipment and consumables guidance.

obtained from offsite. Final phase:

4.16 kV mobile GTG is connected to Train A or Train B Class 1E switchgear. Consumables such as cooling water and GTG fuel are supplied from offsite for long-term coping with the event.

3.0 Core Cooling The first set of strategies necessary to meet the requirements of Order The APR1400 FLEX strategy complies with the Strategies EA-12-049 addresses challenges to core cooling. Core cooling must guidance.

be accomplished in all three phases described in the Order. The Supporting analysis for the operational strategy for purpose of these strategies is to provide a means of cooling the core core cooling were performed using RELAP5/Mod 3.3.

in order to prevent fuel damage. It was shown from the analysis that even in the ELAP concurrent with LUHS, the plant is maintained at safe shutdown state (hot standby, hot shutdown, or cold shutdown, depending on the phase of the APR1400 core cooling FLEX strategy) without fuel damage.

4.0 Spent Fuel The second set of strategies necessary to meet the requirements of The APR1400 FLEX strategy complies with this Pool Cooling Order EA-12-049 addresses challenges to SFP cooling. guidance.

Strategies SFP cooling must be accomplished in all three phases described in Alternate means for supplying both makeup water to the Order. the SFP and spray water are provided from the The purpose of these strategies is to provide alternate means of outside of the auxiliary building. Maximum core cooling the spent fuel in order to prevent fuel damage. Licensees must offload condition was considered in the SFP boiloff consider all loading conditions relevant to their SFP, including a analysis.

maximum core offload.

5.0 Containment The third group of strategies and guidance necessary to meet the Upon loss of all ac power, all containment Function requirements of Order EA 049 addresses challenges to the penetrations are isolated by either using dc power or Strategies containment functions. mobile ac power. Also, for those penetrations Containment functions must be accomplished in all three phases needed to be opened for FLEX strategies, the described in the Order. isolation valves (SI-601, IW-005, IW-006, and CS-1013) can be opened by manual operation.

KEPCO & KHNP 59

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-7 Conformance with JLD-ISG-2012-01, Rev. 0 (4 of 5)

JLD-ISG-2012-01 Rev. 0 APR1400 Section Summary 5.1 Removal of Beyond-design-basis external events such as a prolonged SBO or loss The APR1400 FLEX strategy complies with this Heat from of normal access to the ultimate heat sink could result in a long-term guidance.

Containment loss of containment heat removal. The containment pressure and temperature can be (Pressure The goal of this strategy is to relieve pressure from the containment in maintained below the design basis value since the Control) such an event. only source of energy imparted onto the containment building is the RCP seal leakage, which is 94.64 L/min (25 gpm) / RCP (a total of 378.54 L/min [100 gpm]).

This is well below the mass and energy of the design basis accident, and the containment integrity can be maintained for the BDBEE conditions.

6.0 Programmatic Controls 6.1 Equipment Storage locations chosen for the equipment must provide protection COL applicant is responsible for the FLEX equipment Protection, from external events as necessary to allow the equipment to perform protection, storage, and deployment.

Storage, and its function without loss of capability. In addition, the licensee must Deployment provide a means to bring the equipment to the connection point under those conditions in time to initiate the strategy prior to expiration of the estimated capability to maintain core and spent fuel pool cooling and containment functions in the initial response phase.

Staff Position: NEI 12-06 provides an acceptable method to provide reasonable protection, storage, and deployment of the equipment associated with Order EA-12-049.

KEPCO & KHNP 60

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-7 Conformance with JLD-ISG-2012-01, Rev. 0 (5 of 5)

JLD-ISG-2012-01 Rev. 0 APR1400 Section Summary 6.2 Equipment Staff Position: NEI 12-06 provides an acceptable method to control the COL applicant is responsible for the FLEX equipment Quality quality of equipment associated with Order EA-12-049 with the quality assurance.

following clarifications.

1. Installed structures, systems and components pursuant to 10 CFR 50.63(a) should continue to meet the augmented quality guidelines of Regulatory Guide 1.155, Station Blackout.

2. Development of maintenance and testing programs for the portable equipment responsive to Order EA-12-049, following the guidelines of NEI 12-06 and standard industry processes for ensuring equipment reliability, provides an acceptable method to reasonably assure the equipment will be functional.

3. In the absence of consensus standards specifically developed for these mitigating strategies, a licensees conformance to consensus standards developed for similar emergency uses, such as those of the National Fire Protection Association for fire protection equipment, provides an acceptable method to reasonably assure the equipment will be functional.

7.0 Guidance for Appendix F of NEI 12-06 provides specific guidance for licensees with Not applicable to the APR1400 AP1000 reactors of the AP1000 design on how to satisfy provisions of Order Design EA-12-049, Attachment 3, for the final phase (for sufficient offsite resources to sustain functions indefinitely).

KEPCO & KHNP 61

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-8 Conformance with NEI 12-06, Rev. 0 (1 of 20)

NEI 12-06, Rev. 0 APR1400 Section Summary 1.0 Introduction This guide outlines the process to be used by licensees, Not a requirement.

Construction Permit (CP) holders, and Combined License (COL) holders to define and deploy strategies that will enhance their ability to cope with conditions resulting from beyond-design-basis external events.

1.1 Background 1.2 Purpose 1.3 Objectives and The objective of FLEX is to establish an indefinite coping The APR1400 FLEX strategy complies with the Guiding capability to prevent damage to the fuel in the reactor and spent guidance.

Principles fuel pools and to maintain the containment function by using The APR1400 FLEX strategy to cope with installed equipment, onsite portable equipment, and pre-staged simultaneous occurrence of ELAP and LUHS has offsite resources. been developed to establish an indefinite coping This capability will address both an ELAP (i.e., loss of offsite capability to prevent damage to the fuel in the reactor power, emergency diesel generators but not the loss of ac power and spent fuel pools and to maintain the containment to buses fed by station batteries through inverters) and a LUHS function by using installed equipment, onsite FLEX (loss of ultimate heat sink) which could arise following external equipment, and offsite resources.

events that are within the existing design basis with additional failures and conditions that could arise from a beyond-design- It follows the three-phase approach as guided by NEI-basis external event. Since the beyond-design-basis regime is 12-06. The three-phase operations consist of an initial essentially unlimited, plant features and insights from beyond- response phase using installed equipment, a transition design-basis evaluations are used, where feasible, to inform phase using FLEX equipment and consumables to coping strategies. provide core cooling, and a third phase of indefinite The underlying strategies for coping with these conditions involve sustainment of these functions using offsite resources.

a three-phase approach:

1) Initially cope by relying on installed plant equipment. The duration of phase 1 and 2 has been justified by

2) Transition from installed plant equipment to onsite FLEX support analysis considering the onsite availability of equipment. equipment, onsite resources.

3) Obtain additional capability and redundancy from offsite equipment until power, water, and coolant injection systems are In the final phase, the 4.16 kV mobile GTG is brought restored or commissioned. to the site and used to restore the safety system such Plant-specific analyses will determine the duration of each phase. as SCS. Consumables such as cooling water and Recovery of the damaged plant is beyond the scope of FLEX GTG fuel are assumed to be supplied from offsite for capabilities as the specific actions and capabilities will be a long-term coping with the event.

function of the specific condition of the plant and these conditions cannot be known in advance.

KEPCO & KHNP 62

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-8 Conformance with NEI 12-06, Rev. 0 (2 of 20)

NEI 12-06, Rev. 0 APR1400 Section Summary To the extent practical, generic thermal hydraulic analyses will be Site-specific procedure to bring offsite resources will developed to support plant-specific decision-making. Justification be prepared by COL applicants.

for the duration of each phase will address the onsite availability of equipment, the resources necessary to deploy the equipment consistent with the required timeline, anticipated site conditions following the beyond-design-basis external event, and the ability of the local infrastructure to enable delivery of equipment and resources from offsite.

1.4 Relationship to Other Tier 1 Requirements 1.5 Applicability This guidance document is applicable to operating reactors, This guidance is applied to the APR1400 design construction permit holders, and COL holders and addresses the certification except for the provisions for which COL development of mitigation strategies for beyond-design-basis applicants are responsible.

external events.

2.0 Overview of FLEX strategies will be determined based on two criteria. Each The COL applicants are responsible to finalize the Implementation plant will establish the ability to cope with the baseline conditions FLEX protection and deployment strategies in Process for a simultaneous ELAP and LUHS event. consideration of the site-specific external hazards.

Each plant would then evaluate the FLEX protection and deployment strategies in consideration of the challenges of the external hazards applicable to the site. Depending on the challenge presented, the approach and specific implementation strategy may vary.

Each plant and site has unique features and for this reason, the implementation of FLEX capabilities will be site-specific. This guideline is organized around the site assessment process shown in Figure 2-1. The guidance is provided to outline the steps, considerations, and ultimate FLEX strategies that are to be provided for each site.

2.1 Establish The first step of FLEX capability development is the establishment The APR1400 FLEX strategy complies with the Baseline Coping of the baseline coping capability to address a simultaneous ELAP guidance.

Capability and LUHS event. In general, the baseline coping capability is established based on an assumed set of boundary conditions that arise from a beyond-design-basis external event. Each plant will establish the ability to cope for KEPCO & KHNP 63

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-8 Conformance with NEI 12-06, Rev. 0 (3 of 20)

NEI 12-06, Rev. 0 APR1400 Section Summary these baseline conditions using a combination of installed, temporary, and offsite equipment. These capabilities will also improve the ability of each plant to respond to other causes of a simultaneous ELAP and LUHS not specifically the result of an external event.

2.2 Determine This step of the site assessment process involves the evaluation COL applicants are responsible to conduct the Applicable of the external hazards that are considered credible to a evaluation of the site-specific external hazards in External particular site. For the purposes of this assessment, external accordance with the guidance.

Hazards hazards have been grouped into five classes to help further focus the effort:

seismic events

external flooding

storms such as hurricanes, high winds, and tornadoes

extreme snow, ice, and cold

extreme heat Each plant will evaluate the applicability of these hazards and, where applicable, address the implementation considerations associated with each. These considerations include:

protection of FLEX equipment

deployment of FLEX equipment

procedural interfaces

usation of offsite resources 2.3 Define This step involves the consideration of the hazards that are COL applicants are responsible to define the site-Site-Specific applicable to the site, in order to establish the best overall specific FLEX strategies.

FLEX Strategies strategy for the deployment of FLEX capabilities for beyond-design-basis conditions.

Considering the external hazards applicable to the site, the FLEX mitigation equipment should be stored in a location or locations such that it is reasonably protected such that no one external event can reasonably fail the site FLEX capability.

Reasonable protection can be provided for example, through provision of multiple sets of portable onsite equipment stored in diverse locations or through storage in structures designed to reasonably protect from applicable external events.

The process for defining the full extent of the FLEX coping capability is described in Section 10.

KEPCO & KHNP 64

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-8 Conformance with NEI 12-06, Rev. 0 (4 of 20)

NEI 12-06, Rev. 0 APR1400 Section Summary 2.4 Programmatic The programmatic controls for implementation of FLEX include: COL applicants are responsible to establish the Controls

quality attributes programmatic controls for implementation of FLEX

equipment design and to coordinate them within the site procedural

equipment storage framework.

procedure guidance

maintenance and testing

training

staffing

configuration control Procedures and guidance to support deployment and implementation including interfaces to EOPs, special event procedures, abnormal event procedures, and system operating procedures, will be coordinated within the site procedural framework.

2.5 Synchronization The timely provision of effective offsite resources will need to be COL applicants are responsible to arrange the offsite with Offsite coordinated by the site and will depend on the plant-specific equipment and resources that required for the offsite Resources analysis and strategies for coping with the effects of the beyond- phase based on the APR1400 FLEX strategies.

design-basis external event. Arrangements will need to be established by each site for the offsite equipment and resources that will be required for the offsite phase. The offsite response interfaces for FLEX capabilities are described in Section 12.

3.2 Performance See below.

Attributes 3.2.1 General Criteria See below.

and Baseline Assumptions 3.2.1.1 General Criteria Procedures and equipment relied upon should ensure that The APR1400 FLEX strategy complies with the satisfactory performance of necessary fuel cooling and guidance in terms of the criteria "maintain fuel containment functions are maintained. A simultaneous ELAP cooling," "maintain containment function," and "keep and LUHS challenges both core cooling and spent fuel pool fuel in the spent fuel pool covered." However, the cooling due to interruption of normal ac powered system criterion "no fuel failure" is used instead of the operations. guidance "keep the fuel in the reactor covered,"

For a PWR, an additional requirement is to keep the fuel in the because the mixture level in the reactor is not clearly reactor covered. For both PWRs and BWRs, the requirement is defined in the supporting analysis using RELAP5.

to keep fuel in the spent fuel pool covered. Instead of calculating the mixture level, the core KEPCO & KHNP 65

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-8 Conformance with NEI 12-06, Rev. 0 (5 of 20)

NEI 12-06, Rev. 0 APR1400 Section Summary The conditions considered herein are beyond-design-basis. cooling capability is ensured by the fuel heat-up Consequently, it is not possible to bind all essential inputs to analysis which directly calculates the fuel these evaluations. temperature.

3.2.1.2 Initial Plant (1) Prior to the event the reactor has been operating at 100 The APR1400 FLEX strategy complies with the Conditions percent rated thermal power for at least 100 days or has just guidance.

been shut down from such a power history as required by plant procedures in advance of the impending event.

(2) At the time of the postulated event, the reactor and supporting systems are within normal operating ranges for pressure, temperature, and water level for the appropriate plant condition. All plant equipment is either normally operating or available from the standby state as described in the plant design and licensing basis.

3.2.1.3 Initial Conditions (1) No specific initiating event is used. The initial condition is The APR1400 FLEX strategy complies with the assumed to be a loss of offsite power (LOOP) at a plant site guidance.

resulting from an external event that affects the offsite power system either throughout the grid or at the plant with no prospect for recovery of offsite power for an extended period. The LOOP is assumed to affect all units at a plant site.

(2) All installed sources of emergency onsite ac power and SBO Alternate ac power sources are assumed to be not available and not imminently recoverable.

(3) Cooling and makeup water inventories contained in systems or structures with designs that are robust with respect to seismic events, floods, and high winds, and associated missiles are available.

(4) Normal access to the ultimate heat sink is lost, but the water inventory in the UHS remains available and robust piping connecting the UHS to plant systems remains intact. The motive force for UHS flow, i.e., pumps, is assumed to be lost with no prospect for recovery.

(5) Fuel for FLEX equipment stored in structures with designs which are robust with respect to seismic events, floods and high winds and associated missiles, remains available.

KEPCO & KHNP 66

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-8 Conformance with NEI 12-06, Rev. 0 (6 of 20)

NEI 12-06, Rev. 0 APR1400 Section Summary (6) Permanent plant equipment that is contained in structures with designs that are robust with respect to seismic events, floods, and high winds, and associated missiles, are available.

(7) Other equipment, such as portable ac power sources, portable back up dc power supplies, spare batteries, and equipment for 50.54(hh)(2), may be used provided it is reasonably protected from the applicable external hazards per Sections 5 through 9 and Section 11.3 of this guidance and has predetermined hookup strategies with appropriate procedures/guidance and the equipment is stored in a relative close vicinity of the site.

(8) Installed electrical distribution system, including inverters and battery chargers, remain available provided they are protected consistent with current station design.

(9) No additional events or failures are assumed to occur immediately prior to or during the event, including security events.

(10) Reliance on the fire protection system ring header as a water source is acceptable only if the header meets the criteria to be considered robust with respect to seismic events, floods, and high winds, and associated missiles.

3.2.1.4 Reactor Additional boundary conditions: The APR1400 FLEX strategy complies with the Transient (1) Following the loss of all ac power, the reactor automatically guidance.

trips and all rods are inserted.

(2) The main steam system valves (such as main steam isolation valves, turbine stops, atmospheric dumps, etc.), necessary to maintain decay heat removal functions operate as designed.

(3) Safety/Relief Valves (S/RVs) or Power Operated Relief Valves (PORVs) initially operate in a normal manner if conditions in the RCS so require. Normal valve reseating is also assumed.

(4) No independent failures, other than those causing the ELAP/LUHS event, are assumed to occur in the course of the transient.

KEPCO & KHNP 67

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-8 Conformance with NEI 12-06, Rev. 0 (7 of 20)

NEI 12-06, Rev. 0 APR1400 Section Summary 3.2.1.5 Reactor Coolant Sources of expected PWR reactor coolant inventory loss include: The APR1400 FLEX strategy complies with the Inventory Loss (1) normal system leakage guidance.

(2) losses from letdown unless automatically isolated or until During normal operation, there is the controlled bleed-isolation is procedurally directed off of 12.11 L/min (3.2 gpm) through each RCP, which (3) losses due to reactor coolant pump seal leakage (rate is is compensated by charging flow.

dependent on the RCP seal design) RCP seal leakage might progress from the controlled bleed-off of 12.11 L/min (3.2 gpm) per RCP to around 2

75.71 L/min (20 gpm) per RCP at 158.19 kg/cm A (2,250 psia) after 30 minutes.

Other normal system leakages, i.e., identified leakage of 10 gpm and unidentified leakage of 0.5 gpm, are allowed in APR1400 technical specification. Normal letdown flow is 302.83 L/min (80 gpm), but letdown isolation valve is designed to close at setpoint of PZR low pressure. The letdown isolation valve could be also closed by operator action within 30 minutes following the event.

In the support analysis for the APR1400 FLEX strategy, the seal leakage from each RCP is assumed to be 94.64 L/min (25 gpm) from the beginning of the event.

Therefore, the assumption of seal leakage is conservatively determined to include all of system leakages considered above.

3.2.1.6 SFP Conditions Initial conditions: The APR1400 FLEX strategy complies with the (1) All boundaries of the SFP are intact, including the liner, gates, guidance.

transfer canals, etc.

(2) Although sloshing may occur during a seismic event, the initial loss of SFP inventory does not preclude access to the refueling deck around the pool.

(3) SFP cooling system is intact, including attached piping.

(4) SFP heat load assumes the maximum design basis heat load for the site.

3.2.1.7 Event Response Event response actions follow the command and control of the The APR1400 FLEX strategy complies with the Actions existing procedures and guidance based on the underlying guidance.

symptoms that result from the event.

KEPCO & KHNP 68

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-8 Conformance with NEI 12-06, Rev. 0 (8 of 20)

NEI 12-06, Rev. 0 APR1400 Section Summary The priority for the plant response is to utilize systems or equipment that provides the highest probability for success.

Other site impacts as a result of the event would be addressed according to plant priorities and resource availability.

The FLEX strategy relies upon the following principles:

1) Initially cope by relying on installed plant equipment.

2) Transition from installed plant equipment to onsite FLEX equipment.

3) Obtain additional capability and redundancy from offsite resources until power, water, and coolant injection systems are restored or commissioned.

4) Response actions will be prioritized based on available equipment, resources, and time constraints. The initial coping response actions can be performed by available site personnel post-event.

5) Transition from installed plant equipment to onsite FLEX equipment may involve onsite, offsite, or recalled personnel as justified by plant-specific evaluation.

6) Strategies that have a time constraint to be successful should be identified and a basis provided that the time can reasonably be met.

3.2.1.8 Effects of Loss The effects of loss of HVAC in an extended loss of ac power The APR1400 FLEX strategy complies with the of event can be addressed consistent with NUMARC 87-00 [Ref. 8] guidance.

Ventilations or by plant-specific thermal hydraulic calculations, e.g., GOTHIC calculations.

3.2.1.9 Personnel Areas requiring personnel access should be evaluated to ensure The APR1400 FLEX strategy complies with the Accessibility that conditions will support the actions required by the plant- guidance.

specific strategy for responding to the event. The connections for primary and secondary FLEX pumps, and mobile GTGs, are provided on the outside of the exterior wall of the auxiliary building, thereby providing reasonable assurance of the accessibility of personnel and equipment.

3.2.1.10 Instrumentation Actions specified in plant procedures/guidance for loss of ac The APR1400 FLEX strategy complies with the and Controls power are predicated on use of instrumentation and controls guidance.

powered by station batteries. In order to extend battery life, a KEPCO & KHNP 69

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-8 Conformance with NEI 12-06, Rev. 0 (9 of 20)

NEI 12-06, Rev. 0 APR1400 Section Summary minimum set of parameters necessary to support strategy Operator actions defined in the coping strategy for a implementation should be defined. The parameters selected must simultaneous ELAP and LUHS are predicated on use be able to demonstrate the success of the strategies at of instrumentation and controls powered by the dc maintaining the key safety functions as well as indicate imminent battery, which is available for 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months without load or actual core damage to facilitate a decision to manage the shedding and 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months with load shedding started at 8 response to the event within the Emergency Operating hours.

Procedures and FLEX Support Guidelines or within the SAMGs. The essential parameters for operator actions include:

Typically, these parameters would include the following:

SG level

SG Level

SG pressure

SG Pressure

PZR pressure

RCS Pressure

Hot leg temperature

RCS Temperature

Cold leg temperature

Containment Pressure

Containment pressure

SFP Level

SFP level 3.2.1.11 Containment It is assumed that the containment isolation actions delineated in The APR1400 FLEX strategy complies with the Isolation Valves current station blackout coping capabilities is sufficient. guidance.

See response to Section 5.0 of Table 5-8.

3.2.2 Minimum Each site should establish the minimum coping capabilities COL applicants are responsible to develop Baseline consistent with unit-specific evaluation of the potential impacts plant-specific procedures based on the APR1400 Capabilities and responses to an ELAP and LUHS. In general, this coping can FLEX strategy, considering the guidance in NEI 12-06.

be thought of as occurring in three phases:

Phase 1: Cope relying on installed plant equipment.

Phase 2: Transition from installed plant equipment to onsite FLEX equipment.

Phase 3: Obtain additional capability and redundancy from offsite equipment until power, water, and coolant injection systems are restored or commissioned.

In order to support the objective of an indefinite coping capability, each plant will be expected to establish capabilities consistent with Table 3-2 (PWRs).

The following guidelines are provided to support the development of guidance to cope with the existing set of plant operating procedures/guidance:

KEPCO & KHNP 70

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-8 Conformance with NEI 12-06, Rev. 0 (10 of 20)

NEI 12-06, Rev. 0 APR1400 Section Summary (1) Plant procedures/guidance should identify site-specific actions necessary to restore ac power to essential loads. If an Alternate ac (AAC) power source is available it should be started as soon as possible. If not, actions should be taken to secure existing equipment alignments and provide an alternate power source as soon as possible based on relative plant priorities.

(2) Plant procedures/guidance should recognize the importance of AFW/HPCI/RCIC/IC during the early stages of the event and direct the operators to invest appropriate attention to assuring its initiation and continued, reliable operation throughout the transient since this provides reasonable assurance decay heat removal.

(3) Plant procedures/guidance should specify actions necessary to assure that equipment functionality can be maintained (including support systems or alternate method) in an ELAP/LUHS or can perform without ac power or normal access to the UHS.

(4) Plant procedures/guidance should identify the sources of potential reactor inventory loss, and specify actions to prevent or limit significant loss.

(5) Plant procedures/guidance should ensure that a flow path is promptly established for makeup flow to the steam generator/nuclear boiler and identify backup water sources in order of intended use. Additionally, plant procedures/guidance should specify clear criteria for transferring to the next preferred source of water.

(6) Plant procedures/guidance should identify loads that need to be stripped from the plant dc buses (both Class 1E and non-Class 1E) for the purpose of conserving dc power.

(7) Plant procedures/guidance should specify actions to permit appropriate containment isolation and safe shutdown valve operations while ac power is unavailable.

KEPCO & KHNP 71

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-8 Conformance with NEI 12-06, Rev. 0 (11 of 20)

NEI 12-06, Rev. 0 APR1400 Section Summary (8) Plant procedures/guidance should identify the portable lighting (e.g., flashlights or headlamps) and communications systems necessary for ingress and egress to plant areas required for deployment of FLEX strategies.

(9) Plant procedures/guidance should consider the effects of ac power loss on area access, as well as the need to gain entry to the Protected Area and internal locked areas where remote equipment operation is necessary.

(10) Plant procedures/guidance should consider loss of ventilation effects on specific energized equipment necessary for shutdown (e.g., those containing internal electrical power supplies or other local heat sources that may be energized or present in an ELAP.

(11) Plant procedures/guidance should consider accessibility requirements at locations where operators will be required to perform local manual operations.

(12) Plant procedures/guidance should consider loss of heat tracing effects for equipment required to cope with an ELAP.

Alternate steps, if needed, should be identified to supplement planned action.

(13) Use of portable equipment, e.g., portable power supplies, portable pumps, etc., can extend plant coping capability. The procedures/guidance for implementation of these portable systems should address the transitions from installed sources to portable are available as well as to address delivery capabilities.

(14) Procedures/guidance should address the appropriate monitoring and makeup options to the SFP.

3.3 Consideration in Once the analysis determines the equipment requirements for COL applicants are responsible to comply with the Utilizing Off-Site extended coping, the licensee should obtain the required onsite guidance.

Resources equipment and ensure appropriate arrangements are in place to obtain the necessary offsite equipment including its deployment at the site in the time required by the analysis.

KEPCO & KHNP 72

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-8 Conformance with NEI 12-06, Rev. 0 (12 of 20)

NEI 12-06, Rev. 0 APR1400 Section Summary The site will need to identify staging area(s) for receipt of the equipment and a means to transport the offsite equipment to the deployment location.

It is expected that the licensee will ensure the offsite resource organization will be able to provide the resources that will be necessary to support the extended coping duration. A list of possible offsite equipment is provided in Section 12.

In addition, the licensee will need to ensure standard connectors for electrical and mechanical equipment compatible with the site connections are provided.

4 STEP 2: See below.

Determine Applicable Extreme External Hazards 5 STEP 2A: The FLEX deployment strategy will address seismic hazards at all COL applicant is responsible for the site-specific Assess Seismic sites. seismic hazards assessment.

Impact 6 STEP 2B: The potential challenge presented by external flooding is very COL applicant is responsible for the site-specific Assess External site-specific and is a function of the site layout, plant design, and flooding assessment.

Flooding Impact potential external flooding hazards present. Typically, plant design bases address the following hazards:

local intense precipitation

flooding from nearby rivers, lakes, and reservoirs

high tides

seiche

hurricane and storm surge

tsunami events.

7 STEP 2C: The evaluation of high wind-induced challenges has three parts. COL applicant is responsible for the site-specific high-Assess Impact of The first part is determining whether the site is potentially wind assessment.

Severe Storms susceptible to different high wind conditions. The second part is with High Winds the characterization of the applicable high wind threat. The third part is the application of the high wind threat characterization to the protection and deployment of FLEX strategies.

KEPCO & KHNP 73

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-8 Conformance with NEI 12-06, Rev. 0 (13 of 20)

NEI 12-06, Rev. 0 APR1400 Section Summary 8 STEP 2D: All sites should consider the temperature ranges and weather COL applicant is responsible for the site-specific snow, Assess Impact of conditions for their site in storing and deploying their FLEX ice, and extreme cold condition assessment.

Snow, Ice and equipment.

Extreme Cold 9 STEP 2E: All sites will address high temperatures. COL applicant is responsible for the site-specific high-Assess Impact of temperature assessment.

High Temperatures 10 STEP 3: Define This step involves the consideration of the aggregate set of onsite COL applicant is responsible for the site-specific Site-Specific and offsite resource considerations for the hazards that are aggregated assessment of the FLEX capabilities.

FLEX applicable to the site. That is, the site should aggregate all of the Capabilities considerations related to:

protection of FLEX equipment

deployment of FLEX equipment

procedural interfaces

usation of offsite resources In order to establish the best overall strategy for the storage and deployment of FLEX capabilities over a broad set of beyond-design-basis conditions an aggregated assessment is needed of the site-specific considerations identified for the applicable hazards.

11 Programmatic Controls 11.1 Quality Attributes Equipment associated with these strategies will be procured as COL applicants are responsible to determine quality commercial equipment with design, storage, maintenance, attributes of FLEX equipment.

testing, and configuration control as outlined in this section. If the equipment is credited for other functions (e.g., fire protection),

then the quality attributes of the other functions apply.

11.2 Equipment 1. Design requirements and supporting analysis should be Requirement for major design parameters of FLEX Design developed for portable equipment that directly performs a FLEX equipment such as capacity and voltage of 480 V mitigation strategy for core, containment, and SFP that provides mobile GTGs, and head and flow of FLEX pumps have the inputs, assumptions, and documented analysis that the been determined, based on the support analysis.

mitigation strategy and support equipment will perform as KEPCO & KHNP 74

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-8 Conformance with NEI 12-06, Rev. 0 (14 of 20)

NEI 12-06, Rev. 0 APR1400 Section Summary intended. When specifying portable equipment, the capacities COL applicants are responsible to determine site-should ensure that the strategy can be effective over a range of specific design requirement for FLEX equipment that plant and environmental conditions. This documentation should directly performs the FLEX mitigation.

be auditable, consistent with generally accepted engineering principles and practices, and controlled within the configuration document control system.

2. Portable towable equipment that is designed for over the road transport typically used in construction/remote sites are deemed sufficiently rugged to function following a BDB seismic event.

3. Note that the functionality of the equipment may be outside the manufacturers specifications if justified in a documented engineering evaluation.

4. It is desirable for diverse mitigation equipment to be commonly available (e.g., commercial equipment) such that parts and replacements can be readily obtained.

11.3 Equipment 1. Detailed guidance for selecting suitable storage locations that COL applicant is responsible for equipment storage.

Storage provide reasonable protection during specific external events is provided in Sections 5 through 9.

2. A technical basis should be developed for equipment storage for portable equipment that directly performs a FLEX mitigation strategy for core, containment, and SFP that provides the inputs, assumptions, and documented basis that the mitigation strategy and support equipment will be reasonably protected from applicable external events such that the equipment could be operated in place, if applicable, or moved to its deployment locations. This basis should be auditable, consistent with generally accepted engineering principles, and controlled within the configuration document control system.

3. FLEX mitigation equipment should be stored in a location or locations informed by evaluations performed per Sections 5 through 9 such that no one external event can reasonably fail the site FLEX capability.

KEPCO & KHNP 75

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-8 Conformance with NEI 12-06, Rev. 0 (15 of 20)

NEI 12-06, Rev. 0 APR1400 Section Summary

4. Different FLEX equipment can be credited for independent events.

5. Consideration should be given to the transport from the storage area following the external event recognizing that external events can result in obstacles restricting normal pathways for movement.

6. If portable FLEX equipment is pre-staged such that it minimizes the time delay and burden of hook-up following an external event, then the equipment should be evaluated to not have an adverse effect on existing SSCs and the primary connection point should be as close to the intended point of supply as possible, e.g., a staged power supply to recharge batteries should be connected as close to the battery charger as practicable to maintain diversity and minimize the reliance on other installed equipment.

7. FLEX equipment should be stored and maintained in a manner that is consistent with assuring that it does not degrade over long periods of storage and that it is accessible for periodic maintenance and testing.

8. If 50.54(hh)(2) equipment is credited in the FLEX mitigating strategies, it should meet the above storage requirements in addition to the 50.54(hh)(2) requirements.

9. If debris removal equipment is needed, it should be reasonably protected from the applicable external events such that it is likely to remain functional and deployable to clear obstructions from the pathway between the FLEX equipments storage location and its deployment location(s).

10. Deployment of the FLEX equipment or debris removal equipment from storage locations should not depend on offsite power or onsite emergency ac power (e.g., to operate roll up doors, lifts, elevators, etc.).

11.4 Procedure Guidance KEPCO & KHNP 76

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-8 Conformance with NEI 12-06, Rev. 0 (16 of 20)

NEI 12-06, Rev. 0 APR1400 Section Summary 11.4.1 Objectives The purpose of this section is to describe the procedural Not a requirement approach for the implementation of diverse and flexible (FLEX) strategies. This approach includes appropriate interfaces between the various accident mitigation procedures so that overall strategies are coherent and comprehensive. This approach is intended to provide guidance for responding to BDBEE events while minimizing the need for invoking 50.54 (x).

11.4.2 Operating 1. The existing hierarchy for operating plant procedures remains COL applicants are responsible to organize operating Procedure relatively unchanged with the following exceptions: procedure hierarchy.

Hierarchy a. A new group of FSG for implementation of FLEX strategies will be created.

b. Existing AOPs and EOPs will be revised to the extent necessary to include appropriate portions or reference to FSG.

2. Where FLEX strategies rely on permanently installed equipment, changes may be required to AOPs and EOPs.

3. Transition from the current procedure structure to the modified procedure structure that incorporates the FLEX strategies is illustrated in Figure 11-1.

11.4.3 Development The inability to predict actual plant conditions that require the use COL applicants are responsible to develop EOP and Guidance for of FLEX equipment makes it impossible to provide specific FSG that can be employed for a variety of conditions.

FSGs procedural guidance. As such, the FSG will provide guidance that can be employed for a variety of conditions.

11.4.4 Regulatory NEI 96-07, revision 1, and NEI 97-04, revision 1 should be used COL applicants are responsible to develop EOP and Screening/ to evaluate the changes to existing procedures as well as to the FSG.

Evaluations FSG to determine the need for prior NRC approval. Changes to procedures (EOPs or FSGs) that perform actions in response events that exceed a site's design basis should, per the guidance and examples provided in NEI 96-07, Rev. 1, screen out.

Therefore, procedure steps which recognize the beyond-design-basis ELAP/LUHS has occurred and which direct actions to ensure core cooling, SFP cooling, or containment integrity should not require prior NRC approval.

KEPCO & KHNP 77

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-8 Conformance with NEI 12-06, Rev. 0 (17 of 20)

NEI 12-06, Rev. 0 APR1400 Section Summary 11.5 Maintenance 1. FLEX mitigation equipment should be initially tested or other COL applicants are responsible for maintenance and and Testing reasonable means used to verify performance conforms to the test of FLEX equipment.

limiting FLEX requirements. Validation of source manufacturer quality is not required.

2. Portable equipment that directly performs a FLEX mitigation strategy for the core, containment, or SFP should be subject to maintenance and testing guidance provided in INPO AP 913, Equipment Reliability Process, to verify proper function. The maintenance program should ensure that the FLEX equipment reliability is being achieved. Standard industry templates (e.g.,

EPRI) and associated bases will be developed to define specific maintenance and testing.

3. The unavailability of equipment and applicable connections that directly performs a FLEX mitigation strategy for core, containment, and SFP should be managed such that risk to mitigating strategy capability is minimized.

11.6 Training 1. Programs and controls should be established to assure COL applicants are responsible for training program personnel proficiency in the mitigation of beyond-design-basis and controls.

events is developed and maintained. These programs and controls should be implemented in accordance with an accepted training process.

2. Periodic training should be provided to site emergency response leaders on beyond-design-basis emergency response strategies and implementing guidelines. Operator training for beyond-design-basis event accident mitigation should not be given undue weight in comparison with other training requirements. The testing/evaluation of Operator knowledge and skills in this area should be similarly weighted.

3. Personnel assigned to direct the execution of mitigation strategies for beyond-design-basis events will receive necessary training to ensure familiarity with the associated tasks, considering available job aids, instructions, and mitigating strategy time constraints.

KEPCO & KHNP 78

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-8 Conformance with NEI 12-06, Rev. 0 (18 of 20)

NEI 12-06, Rev. 0 APR1400 Section Summary

4. ANSI/ANS 3.5, Nuclear Power Plant Simulators for use in Operator Training certification of simulator fidelity (if used) is considered to be sufficient for the initial stages of the beyond-design-basis external event scenario until the current capability of the simulator model is exceeded. Full scope simulator models will not be upgraded to accommodate FLEX training or drills.

5. Where appropriate, the integrated FLEX drills should be organized on a team or crew basis and conducted periodically; with all time-sensitive actions to be evaluated over a period of not more than eight years. It is not the intent to connect to or operate permanently installed equipment during these drills and demonstrations.

11.7 Staffing 1. Onsite staff are at site administrative minimum shift staffing COL applicants are responsible for onsite staffing.

levels, (minimum staffing may include additional staffing that is procedurally brought onsite in advance of a predicted external event, e.g., hurricane).

2. No independent, concurrent events, e.g., no active security threat, and

3. All personnel onsite are available to support site response.

11.8 Configuration 1. The FLEX strategies and basis will be maintained in an overall COL applicants are responsible for the plant Control program document. This program document will also contain a configuration control.

historical record of previous strategies and the basis for changes.

The document will also contain the basis for the ongoing maintenance and testing programs chosen for the FLEX equipment.

2. Existing plant configuration control procedures will be modified to ensure that changes to the plant design, physical plant layout, roads, buildings, and miscellaneous structures will not adversely impact the approved FLEX strategies.

3. Changes to FLEX strategies may be made without prior NRC approval.

12 Off-Site Resources KEPCO & KHNP 79

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-8 Conformance with NEI 12-06, Rev. 0 (19 of 20)

NEI 12-06, Rev. 0 APR1400 Section Summary 12.1 Synchronization Arrangements will need to be established by each site addressing COL applicants are responsible to establish with Offsite the scope of equipment that will be required for the offsite phase, arrangement of offsite resources.

Resources as well as the maintenance and delivery provisions for such equipment.

12.2 Minimum Each site will establish a means to ensure the necessary COL applicants are responsible to establish a means Capabilities of resources will be available from offsite. Considerations that to ensure the necessary resources are available from Offsite should be included in establishing this capability include: offsite.

Resources 1) A capability to obtain equipment and commodities to sustain and backup the sites coping strategies.

2) Offsite equipment procurement, maintenance, testing, calibration, storage, and control.

3) A provision to inspect and audit the contractual agreements to reasonably assure the capabilities to deploy the FLEX strategies including unannounced random inspections by the Nuclear Regulatory Commission.

4) Provisions to ensure that no single external event will preclude the capability to supply the needed resources to the plant site.

5) Provisions to ensure that the offsite capability can be maintained for the life of the plant.

6) Provisions to revise the required supplied equipment due to changes in the FLEX strategies or plant equipment or equipment obsolescence.

7) The appropriate standard mechanical and electrical connections need to be specified.

8) Provisions to ensure that the periodic maintenance, periodic maintenance schedule, testing, and calibration of offsite equipment are comparable/ consistent with that of similar onsite FLEX equipment.

9) Provisions to ensure that equipment determined to be unavailable/non-operational during maintenance or testing is either restored to operational status or replaced with appropriate alternate equipment within 90 days.

KEPCO & KHNP 80

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-8 Conformance with NEI 12-06, Rev. 0 (20 of 20)

NEI 12-06, Rev. 0 APR1400 Section Summary

10) Provision to ensure that reasonable supplies of spare parts for the offsite equipment are readily available if needed. The intent of this provision is to reduce the likelihood of extended equipment maintenance (requiring in excess of 90 days for returning the equipment to operational status).

13 Submittal Reporting requirements are established in accordance with NRC NA Guidance Order EA-12-049 Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events.

KEPCO & KHNP 81

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-9 Conformance with NEI 12-06, Rev. 0 - Tables D-1, D-2, and D-3 (1 of 4)

NEI 12-06, Rev. 0 - Tables D-1, D-2, and D-3 APR1400 Safety Function Method Performance Attributes Core Cooling and AFW/EFW

Extend installed coping capability through

Train C/D dc battery is available for 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months Heat procedural enhancements (e.g., load shedding), without load shedding, and 480 V mobile GTG Removal (SG provision of portable battery chargers and other is prepared to charge Train A or Train B dc available) power supplies. battery and to supply ac power to the installed

Objective is to provide extended baseline coping safety components such as ACP.

capability with installed equipment.

ac-independent FLEX pumps provide the

Procedures/guidance to include local manual safety functions such as core cooling and initiation of ac-independent RCS makeup, according to the APR1400 AFW/EFW pumps consistent with NEI 06-12. FLEX strategy.

Core Cooling and Depressurize SG for

Primary and alternate injection points are required

Rated flow of secondary FLEX pump is Heat makeup with portable to establish capability to inject through separate 1173.48 L/min (310 gpm), which is sufficient Removal (SG injection source divisions/trains, i.e., should not have both not only to remove decay heat but also to available) connections in one division/train. restore SG water level.

Makeup paths supply required SGs. *Supporting analysis for FLEX strategy shows

SG makeup rate should exceed decay heat levels that the APR1400 plant has capability for at time of planned deployment in order to support continued core cooling during ELAP (1) restoring SG water level, e.g., 200 gpm. concurrent with LUHS.

Analysis should demonstrate that the guidance and equipment for combined SG depressurization and makeup capability support continued core cooling.

Core Cooling and Sustained source of

Water source sufficient to supply water indefinitely *Onsite water sources such as AFWST and Heat water including consideration of concurrent makeup or RWT provide water to feed SG for Removal (SG spray of SFP approximately 2 weeks. When RWT inventory available) is shared with the SFP cooling water, the water source can feed SG at least for 11 days.

(1)

RCS Inventory Control Low-leakage RCP

Makeup capability to maintain core cooling *The APR1400 RCP adopts a three-stage seal

/ Long-Term seals and/or borated

Sufficient letdown to support required makeup and design, which is similar to CE-KSB pump.

(1)

Subcriticality high-pressure RCS ensure subcriticality *ACP provides RCS with borated water from makeup required IRSWT and BAST, after 480 V mobile GTG is connected.

SIT also provides RCS with borated water, when RCS pressure reduces to the setpoint during cooldown operation.

Primary FLEX pump is also able to make up RCS inventory with borated water in the long KEPCO & KHNP 82

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-9 Conformance with NEI 12-06, Rev. 0 - Tables D-1, D-2, and D-3 (2 of 4)

NEI 12-06, Rev. 0 - Tables D-1, D-2, and D-3 APR1400 Safety Function Method Performance Attributes term.

Core Cooling and All plants provide

Diverse injection points or methods are required to *ACP provides RCS with borated water from Heat Removal (Modes means to provide establish capability to inject through separate IRWST, after 480 V mobile GTG is connected.

5 and 6, SG borated RCS makeup divisions/trains, i.e., should not have both

Rated flow of the primary low-head FLEX (2) unavailable) connections in one division/train. pump is around 2,839 L/min (750 gpm), which

Connection to RCS for makeup should be capable is sufficient for core heat removal.

of flow rates sufficient for simultaneous core heat removal and boron flushing (combined makeup flow (1) exceeding 300 gpm).

Onsite pump (portable or installed) for RCS makeup. This can be the SG makeup pump since both will not be required at same time.

In order to address the requirement for diversity, if repowering of installed charging pumps is used for this function, then either (a) multiple power connection points should be provided to the charging pump, or (b) provide a single power supply connection point for the charging pump and a single connection point for a portable makeup pump.

Source of borated water could be an onsite tank or could be provided by offsite resources.

Key Reactor

Identify instruments to be relied upon, including

Instruments for the following key plant Parameters

SG level control room and field instruments. parameters are available for operators to

SG pressure

Depending on strategy employed, additional monitor plant condition and to carry out

RCS pressure parameters may be required. operator action according to the APR1400

RCS temperature FLEX strategy.

- PZR pressure

- PZR level

- Hot leg temperature

- Cold leg temperature

- SG pressure

- SG level

- Charging flow

- SIT level

- SIT pressure

- Etc.

KEPCO & KHNP 83

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-9 Conformance with NEI 12-06, Rev. 0 - Tables D-1, D-2, and D-3 (3 of 4)

NEI 12-06, Rev. 0 - Tables D-1, D-2, and D-3 APR1400 Safety Function Method Performance Attributes Containment Containment spray Due to the long-term nature of this function, the The emergency containment spray backup Function connection does not need to be a permanent system (ECSBS) is provided for long-term modification. However, if a temporary connection is maintenance of containment function. The necessary, e.g., via valve bonnet, then this should be ECSBS is supplied water by FLEX pump pre-identified. through connections located outside the exterior wall of the auxiliary building.

Key Containment Containment Identify instruments to be relied upon, including The following containment pressure Parameters pressure control room and field instruments. instruments are available in MCR for operators to monitor plant condition and carry out operator action according to the APR1400 FLEX strategy:

- Containment pressure: high alarm indicator

- Containment pressure: high-high alarm indicator Spent Fuel Cooling Makeup with Minimum makeup rate must be capable of exceeding The APR1400 strategy complies with this portable injection boiloff rate for the boundary conditions described in guidance.

source Subsection 3.2.1.6. The hose stations on the operating floor of the (makeup via hoses SFP area can provide the makeup capacity of on refuel floor) 1,893 L/min (500 gpm), which exceeds the maximum boiloff rate (493.2 L/min [130.3 gpm]) .

Makeup with portable Minimum makeup rate must be capable of exceeding The APR1400 strategy complies with this injection source boiloff rate for the boundary conditions described in guidance.

(makeup via Subsection 3.2.1.6. The FLEX makeup capacity (1,893 L/min [500 connection to SFP gpm]) exceeds the maximum boiloff rate cooling piping or other (493.2 L/min [130.3 gpm]).

alternate location)

Makeup with portable Plant-specific strategy should be considered as The rollup door at fuel handling area truck bay injection source (vent needed. in the auxiliary building can be opened to pathway for steam and provide a vent path for steam and condensate condensate from SFP) from SFP.

KEPCO & KHNP 84

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-9 Conformance with NEI 12-06, Rev. 0 - Tables D-1, D-2, and D-3 (4 of 4)

NEI 12-06, Rev. 0 - Tables D-1, D-2, and D-3 APR1400 Safety Function Method Performance Attributes Makeup with portable Minimum of 757 L/min (200 gpm) per unit to the pool The spray makeup of at least 757 L/min (200 injection source (spray or 946 L/min (250 gpm) per unit if overspray occurs gpm) through diverse spray line of 10 cm (4 capability via portable consistent with 10 CFR 50.54(hh)(2). in) is available. Minimum of 757 L/min (200 monitor nozzles from This capability is not required for sites that have gpm) is consistent with the guidance of NEI refueling floor using SFPs that cannot be drained. 06-12.

portable pump).

SFP Parameters SFP level Per EA 12-051 The APR1400 design includes redundant, safety-related wide-range level sensors in SFP that fulfill EA 12-051 order.

Notes:

(1) Subject to generic or plant-specific analysis (2) There may be short periods of time during Modes 5 and 6 when plant configuration may preclude use of this strategy.

KEPCO & KHNP 85

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-10 Conformance with NEI 12-02, Rev. 1 (1 of 11)

NEI 12-02 Rev. 1 APR1400 Section Summary 1.0 Introduction The guidance in this document presents an acceptable method for The APR1400 action plans for guidance in NEI implementing Order EA-12-051, "Issuance of Order to Modify 12-02, Rev. 1 are described in this table.

Licenses with regard to Reliable Spent Fuel Pool Instrumentation."

2.0 Levels Required Monitoring 2.1 Introduction Order EA-12-051 includes requirements as follows: Conformance.

All licensees identified in Attachment 1 to this Order shall have a reliable indication of the water level in associated spent fuel storage Two safety-related level instruments are pools capable of supporting identification of the following pool water provided in the SFP to monitor level (1) to level conditions by trained personnel: support operation of the normal fuel pool cooling (1) level that is adequate to support operation of the normal fuel pool system, (2) to provide substantial radiation cooling system, shielding, and (3) to keep fuel covered and (2) level that is adequate to provide substantial radiation facilitate actions to implement makeup water shielding for a person standing on the spent fuel pool operating addition.

deck, and (3) level where fuel remains covered and actions to implement makeup water addition should no longer be deferred.

2.2 Rational During the events at Fukushima Dai-ichi, responders were NA without reliable instrumentation to determine water level in the spent fuel pool. This led to NRC concerns that the Fukushima Dai-ichi Unit 4 pool might have boiled dry, resulting in significant fuel damage. The events at Fukushima Dai-ichi demonstrated the confusion and misapplication of resources that may result from beyond-design-basis external events when reliable spent fuel pool level instrumentation is not available.

2.3 Wide Range The requirement from this order is for instrumentation that Conformance.

Pool Level covers a wide level range within the spent fuel pool.

Instrumentation The three critical levels that must be monitored in a spent fuel Two safety-related level instruments are pool are described below. provided in the SFP to monitor SFP water level It should be noted that continuous indication from a single from level 1 to level 3 and have a capability of instrument over the entire span from level 1 to level 3 is not continuous indication.

required but could be used. One level instrumentation channel is separated If more than one instrument is used to monitor the entire span, from the other instrument channel.

KEPCO & KHNP 86

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-10 Conformance with NEI 12-02, Rev. 1 (2 of 11)

NEI 12-02 Rev. 1 APR1400 Section Summary that set of instruments constitutes a single channel satisfying either the primary or backup instrument channel requirement (refer to Figure 1).

A visual representation of monitoring levels 1, 2 and 3 and the associated requirements for monitoring between the points are presented in Figure 1.

The minimum requirements apply to the separation distance between level indications and support development of appropriate response procedures. These requirements are separate from the instrument channel design accuracy described in Section 3, which apply to either discrete or to continuous instruments.

2.3.1 Level-1 Level 1 represents the HIGHER of the following two points: Conformance.

Level that is

The level at which reliable suction loss occurs due to adequate to uncovering of the coolant inlet pipe, weir or vacuum breaker

The SFP water level instruments cover Level support (depending on the design), or 1.

operation of

The level at which the water height, assuming saturated

There are siphon breaker holes above the the normal conditions, above the centerline of the cooling pump suction centerline of the cooling pump suction to fuel pool provides the required net positive suction head specified by the provide the required net positive suction head cooling system pump manufacturer or engineering analysis. specified by the pump manufacturer or engineering analysis.

2.3.2 Level-2 Level 2 is based on either of the following: Conformance.

Level that is

10 ft (+/- 1 ft) above the highest point of any fuel rack adequate to seated in the spent fuel pools, or

The SFP water level instruments cover the provide

a designated level that provides adequate radiation shielding Level 2.

substantial to maintain personnel radiological dose levels within acceptable

The instruments cover 3.05 m (10 ft) above the radiation limits while performing local operations in the vicinity of the pool. top of spent fuel rack stored in the SFP to shielding for This level shall be based on either plant-specific or appropriate provide adequate radiation shielding to maintain a person generic shielding calculations, considering the emergency conditions personnel radiological dose levels within standing on that may apply at the time and the scope of necessary local acceptable limits while performing local the SFP operation operations, including installation of portable SFP instrument channel operations in the vicinity of the pool.

deck components.

Designation of this level should not be interpreted to imply that actions to initiate water makeup must be delayed until SFP water levels have reached or are lower than this point.

KEPCO & KHNP 87

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-10 Conformance with NEI 12-02, Rev. 1 (3 of 11)

NEI 12-02 Rev. 1 APR1400 Section Summary 2.3.3 Level-3 Level 3 corresponds nominally (i.e., +/- 1 ft) to the highest Conformance.

Level where point of any fuel rack seated in the spent fuel pool.

fuel remains Level 3 is defined in this manner to provide the maximum range of The SFP water level instruments cover Level 3 covered and information to operators, decision makers and emergency response to provide the maximum range of information to actions to personnel. Designation of this level should not be interpreted to operators, and to provide water coverage over implement makeup imply that actions to initiate water makeup must or should be the spent fuel through timely makeup.

water addition delayed until this level is reached.

should no longer be differed 3.0 Instrumentation design features 3.1 Instruments This instrumentation shall consist of at least one primary and Conformance.

one backup instrument channel.

The spent fuel pool level instruments are Reliable level indication shall be functional (for fixed channels) permanent and fixed channel.

or functional when installed (for portable channels or combination of fixed and portable component channels) during The APR1400 conforms to this requirement but all modes of operation consistent with paragraph 4.3, Testing portable instrument is not used.

and Calibration.

If portable components are used as part of a backup instrument channel then, to limit personnel resources required for deployment, it shall be designed such that it can easily be deployed by a maximum of two trained personnel within 30 minutes at the spent fuel pool (i.e., no more than 1 person-hour).

Portable instrument components must be placed in service in predetermined accessible locations.

However, in anticipation that such predetermined locations may be inaccessible at the time of the event, guidance must be provided to the trained personnel as to how to determine and use alternate locations.

KEPCO & KHNP 88

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-10 Conformance with NEI 12-02, Rev. 1 (4 of 11)

NEI 12-02 Rev. 1 APR1400 Section Summary Portable instrument components must be stored in predetermined accessible locations that will not hinder the ability of trained personnel to install the portable components when needed.

3.2 Arrangement Installation of the SFP instrument channels shall be consistent Conformance.

with the plant-specific SFP design requirements and should not impair normal SFP function. The SFP level instrument channels are arranged to reduce the potential for falling Channel separation should be maintained by locating the debris or missiles affecting both channels of installed sensors in different places in the SFP area. instrumentation and to provide reasonable Provisions for portable instruments should also consider the protection of its function against external need for physical separation. hazards.

Plans for portable instrument use should allow inserting and In the SFP area, cables are routed in operating the sensors and associated equipment in a different seismically mounted rigid metal conduit. Outside part of the SFP from the permanent channel. the pool area, cables are routed in seismically mounted rigid metal conduit, trays, or raceways.

Ideally the portable channel will be able to use multiple (or all)

SFP locations.

Similarly, cabling for power supplies and indications for each channel should be routed separately from cabling for the other channels.

To the extent not otherwise covered in this guidance, the reasonable protection guidance outlined in NEI 12-06 to meet Order EA-12-049 should be used to provide protection for installed and portable channels from external hazards.

At a minimum, cables routed outside structures should be protected in buried conduit and designed to commercial standards for submergence.

3.3 Mounting Consideration shall be given to the maximum seismic ground Conformance.

motion to the design basis of the SFP structure.

KEPCO & KHNP 89

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-10 Conformance with NEI 12-02, Rev. 1 (5 of 11)

NEI 12-02 Rev. 1 APR1400 Section Summary The mounting shall be designed consistent with the highest Instrumentation is designed as Safety Class 3 seismic or safety classification of the SFP. and seismic Category I.

An evaluation of other hardware stored in the SFP shall be conducted to ensure it will not create adverse interaction with the fixed instrument location(s).

The basis for the seismic design for mountings in the SFP shall be the plant seismic design basis at the time of submittal of the Integrated Plan for implementing NRC Order EA-12-051 (See Appendix A-2-2).

3.4 Qualification The instrument channel reliability shall be demonstrated via an See the APR1400 Actions for Qualification of (Guidance) appropriate combination of design, analyses, operating Conditions for shock, vibration and seismic as experience, and/or testing of channel components for the described below.

following sets of parameters, as described in the paragraphs below:

conditions in the area of instrument channel component use for all instrument components,

effects of shock and vibration on instrument channel components used during any applicable event for only installed components, and

seismic effects on instrument channel components used during and following a potential seismic event for only installed components.

Selection of instrument channel components should consider ease and simplicity of design and replacement after the event.

Readily available commercial components shall be considered.

Qualification The temperature, humidity and radiation levels consistent with Conformance.

(Conditions) conditions in the vicinity of the SFP and the area of use considering normal operational, event and post-event The temperature, humidity, and radiation levels conditions for no fewer than seven days post-event or until consistent with conditions in the vicinity of the offsite resources can be deployed by the mitigating strategies SFP and the area of use considering normal resulting from Order EA-12-049 should be considered. operational, event, and post-event conditions for no fewer than 7 days post-event are considered.

KEPCO & KHNP 90

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-10 Conformance with NEI 12-02, Rev. 1 (6 of 11)

NEI 12-02 Rev. 1 APR1400 Section Summary The SFP instrument is designed to operate in borated and non-borated water for expected radiation levels and environments of greater than 100 °C (212 °F) and 100 % humidity.

Qualification For the effects of shock and vibration in the area of instrument Conformance.

(Shock and channel component use after an event for applicable Vibration) components (with the exception of battery chargers and For the effects of shock and vibration in the area replaceable batteries), the following measures are acceptable of instrument channel component use after an to verify that the design and installation is adequate. event for applicable components (with the exception of battery chargers and replaceable Applicable components of the instrument channels are rated by the batteries), either of the methods described in manufacturer (or otherwise tested) for shock and vibration at levels the NEI guidance is used.

commensurate with those of postulated design basis event conditions in the area of instrument channel component use using one or more of the following methods:

instrument channel components use known operating principles, are supplied by manufacturers with commercial quality programs (such as ISO 9001) with shock and vibration requirements included in the purchase specification and/or instrument design, and commercial design and testing for operation in environments where significant shock and vibration loadings are common, such as for portable hand-held devices or transportation applications;

substantial history of operational reliability in environments with significant shock and vibration loading, such as transportation applications; or

use of components inherently resistant to shock and vibration loadings or are seismically reliable such as cables.

Qualification For seismic effects on instrument channel components used Conformance.

(Seismic) after a potential seismic event for only installed components (with the exception of battery chargers and replaceable Instrumentation is designed as Safety Class 3 batteries), the following measures are acceptable to verify that and seismic Category I.

the design and installation is adequate.

KEPCO & KHNP 91

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-10 Conformance with NEI 12-02, Rev. 1 (7 of 11)

NEI 12-02 Rev. 1 APR1400 Section Summary Applicable components of the instrument channels are rated by the Either of the methods described in the NEI 12-manufacturer (or otherwise tested) for seismic effects at levels 02 guidance and NRC staff positions for NEI12-commensurate with those of postulated design basis 02 specified in the JLD-ISG-2012-03 is used for event conditions, in the area of instrument channel component demonstration of seismic adequacy.

use, using one or more of the following methods:

instrument channel components use known operating principles, are supplied by manufacturers with commercial quality programs (such as ISO9001) with seismic requirements included in the purchase specification and/or instrument design, and commercial design and testing for operation in environments where significant seismic effects are common;

substantial history of operational reliability in environments with significant vibration, such as for portable hand-held devices or transportation applications;

demonstration of seismic reliability using methods that predict the equipments performance by

- analysis,

- testing of the equipment under simulated seismic conditions,

-using a combination of test and analysis, or

-the use of experience data.

demonstration that proposed devices are substantially similar in design to models that have been previously tested for seismic effects in excess of the plant design basis at the location where the instrument is to be installed (g-levels and frequency ranges); or

seismic qualification using seismic motion consistent with that of existing design basis loading at the installation location.

Qualification The basis for the seismic qualification for instrument channel Conformance.

(General) components shall be the plant seismic design basis at the time of submittal of the Integrated Plan for implementing NRC Order EA 051 (See Appendix A-2-2).

3.5 Independence If plant ac or dc power sources are used then the power Conformance.

sources shall be from different buses and preferably different KEPCO & KHNP 92

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-10 Conformance with NEI 12-02, Rev. 1 (8 of 11)

NEI 12-02 Rev. 1 APR1400 Section Summary divisions / channels depending on available sources of power. Each SFP water level instrument channel has a different power source and is physically For two (2) permanently mounted (fixed) instruments in the separated to the extent practicable.

pool, they should be separated to the extent practicable considering existing spent fuel pool construction (reference The primary instrument channel is independent Section 3.2). of the backup instrument channel.

Independence is obtained by physical separation of components between channels and the separate use of Class 1E MCC.

3.6 Power The normal electrical power supply for each channel shall be Conformance.

Supplies provided by different sources such that the loss of one of the channels primary power supply will not result in a loss of power Each instrument channel is normally powered supply function to both channels of SFP level instrumentation. from a 120 Vac 60 Hz distribution panel of Class 1E 480 Vac MCC to support continuous All channels of SFP level instrumentation shall provide the capability monitoring of SFP level. On loss of normal 120 of connecting the channel to a source of power (e.g., portable Vac power from the Class 1E 480 V MCC, each generators or replaceable batteries) independent of the normal plant channels internal UPS automatically transfers ac and dc power systems. For fixed channels this alternate instrument power to a dedicated backup battery.

capability shall include the ability to isolate the installed channel from If normal ac power is restored, the UPS its normal power supply or supplies. automatically transfers instrument power back to the normal ac power. The dedicated backup The portable power sources for the portable and installed batteries are sized to be capable of supporting channels shall be stored at separate locations, consistent with continuous monitoring of SFP level for a the reasonable protection requirements associated with NEI minimum of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of operation. This provides 12-06 (Order EA-12-049). adequate time until the onsite or offsite mobile GTG can supply power by mitigating strategies The portable generator or replaceable batteries should be resulting from the ELAP event. Instrument accessible and have sufficient capacity to support reliable accuracy and performance are not affected by instrument channel operation until offsite resources can be restoration of power or restarting the processor.

deployed by the mitigating strategies resulting from Order EA-12-049 If adequate power supply for either an installed or portable level instrument credits intermittent operation, then the provisions shall be made for quickly and reliably taking the channel out of service and restoring it to service.

KEPCO & KHNP 93

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-10 Conformance with NEI 12-02, Rev. 1 (9 of 11)

NEI 12-02 Rev. 1 APR1400 Section Summary 3.7 Accuracy The instrument channels shall maintain their designed accuracy Conformance.

following a power interruption or change in power source without recalibration. Accuracy of the instrument channels considers SFP conditions of saturated water, steam Accuracy should consider SFP conditions, e.g., saturated environment, or concentrated borated water and water, steam environment, or concentrated borated water. is sufficient to allow trained personnel to determine when the actual level exceeds the Additionally, instrument accuracy should be sufficient to allow specified lower level of each indicating range trained personnel to determine when the actual level exceeds (Levels 1, 2, and 3) without conflicting or the specified lower level of each indicating range (levels 1, 2 ambiguous indication.

and 3) without conflicting or ambiguous indication.

3.8 Testing Static or non-active installed (fixed) sensors can be used and Conformance.

should be designed such that testing and /or calibration can be performed in-situ. The APR1400 use a GWR type as the SFP water level instrumentation.

For microprocessor based channels, the instrument channel design shall be capable of testing while mounted in the pool. The SFP water level instrumentation is designed so that testing and/or calibration can Back-up portable channels shall be designed such that be performed in-situ.

calibration does not require the use of any additional test or The APR1400 conforms with the requirement reference equipment at the time of deployment, i.e., but does not use a portable instrument.

plug-and-play type technology.

The COL applicant should develop the station procedures and preventive maintenance tasks to perform required surveillance testing, calibration, backup battery maintenance, functional checks, and visual inspections of the probes.

3.9 Display SFP level indication from the installed channel shall be Conformance.

displayed in the control room, at the alternate shutdown panel, or another appropriate and accessible location (reference NEI The SFP water level is continuously displayed in 12-06). the MCR and in the remote shutdown room (RSR).

An appropriate and accessible location shall include the following characteristics:

KEPCO & KHNP 94

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-10 Conformance with NEI 12-02, Rev. 1 (10 of 11)

NEI 12-02 Rev. 1 APR1400 Section Summary

occupied or promptly accessible to the appropriate plant staff giving appropriate consideration to various drain down scenarios,

outside of the area surrounding the SFP floor, e.g., an appropriate distance from the radiological sources resulting from an event impacting the SFP,

inside a structure providing protection against adverse weather, and

outside of any very high radiation areas or LOCKED HIGH RAD AREA during normal operation.

If multiple display locations beyond the required "appropriate and accessible location" are desired, then the instrument channel shall be designed with the capability to drive the multiple display locations without impacting the primary appropriate and accessible display.

SFP level indication from a portable channel shall be displayed in an accessible location.

4.0 Program Features 4.1 Training The personnel performing functions associated with these SFP level This section is not the scope of the standard instrumentation channels shall be trained to perform the job specific design.

functions necessary for their assigned tasks (maintenance, calibration, surveillance, etc.). SFP instrumentation The personnel performing functions associated should be installed via the normal with these SFP water level instrumentation modification processes. channels shall be trained to perform the job-specific functions necessary for their assigned In either case utilities should use the Systematic Approach to tasks (maintenance, calibration, surveillance, Training (SAT) to identify the population to be trained. etc.) by the COL applicants.

The SAT process should also determine both the initial and continuing elements of the required training.

4.2 Procedures If, at the time of an event or thereafter until the unit is returned This section is not the scope of the standard to normal service, an instrument channel ceases to function, its design.

function must be recovered within a period of time consistent with the emergency conditions that may apply at the time.

KEPCO & KHNP 95

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 5-10 Conformance with NEI 12-02, Rev. 1 (11 of 11)

NEI 12-02 Rev. 1 APR1400 Section Summary If, at the time of an event or thereafter until the unit is returned When an SFP water level instrument channel to normal service, an instrument channel component must be ceases to function, its function must be replaced, it is acceptable to use commercially available recovered within a period of time consistent with components that may or may not meet all of the qualifications the emergency conditions that may apply at the (Section 3.4) to maintain the instrument channel functionality. time. Therefore, the COL applicants shall have a strategy for this.

All licensees shall have a strategy to ensure SFP water level addition is initiated at an appropriate time consistent with the implementation of NEI 12-06.

4.3 Testing and Processes shall be established and maintained for scheduling This section is not the scope of the standard Calibration and implementing necessary testing and calibration of the design.

primary and backup SFP level instrument channels to maintain the instrument channels at the design accuracy. COL applicants are responsible to establish and maintain scheduling and implementing The testing and calibration of the instrumentation shall be necessary testing and calibration of the SFP consistent with vendor recommendations or other documented level instrument channels to maintain them at basis. the design accuracy and incorporate the specific guidance in this section of NEI 12-02, Rev. 1.

Calibration shall be specific to the mounted instrument and the monitor.

Surveillances or testing to validate functionality of an installed instrument channel shall be performed within 60 days of a planned refueling outage considering normal testing scheduling allowances (e.g., 25%).

Additionally, compensatory actions must be taken if the instrumentation channel is not expected to be restored or is not restored within 90 days.

If a single SFP for the purposes of this order is divided by the closure of a normally open gate(s) such that a portion of the SFP containing fuel used for power production within the last five years is no longer able to be monitored by a required SFP instrumentation channel, then the actions described above must be taken for the impacted instrumentation channel.

KEPCO & KHNP 96

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 TS Figure 5-1 Timeline of the APR1400 FLEX Strategy for Full-Power Operation KEPCO & KHNP 97

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 120.0 100.0 No Spray 80.0 Acceptance Criterion 4.22 kg/cm2 (60 psig)

Pressure (psia) 60.0 40.0 20.0 Spray at the design pressure 0.0 0 2 4 6 8 10 12 14 16 18 20 Time (days)

Figure 5-2 Containment Pressure for Full Power (RCP Seal LOCA)

KEPCO & KHNP 98

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 400.0 EQ Temperature Limit 350.0 182 oC (360 oF)

No Spray 300.0 Temperature (oF) 250.0 200.0 Spray at the design pressure 150.0 100.0 0 2 4 6 8 10 12 14 16 18 20 Time (days)

Figure 5-3 Containment Temperature for Full Power (RCP Seal LOCA)

KEPCO & KHNP 99

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 6.0 DESIGN FEATURES AND PROGRAMS TO ADDRESS BDBEE This chapter compiles design enhancements and programs that are incorporated into the APR1400 design to cope with the lessons learned from the accidents at TEPCOs Fukushima Dai-ichi Nuclear Power Station, and satisfy the requirements/recommendations issued after the disaster by the U.S. NRC.

Design features and program descriptions, design basis, and compliance with NRC recommendations are described herein.

6.1 Overall Description The following is the overall description:

- Fukushima issues are described in DCD Chapter 19.3.

- Compliance with NRC guidance is described in DCD Tier 2, Section 1.9.

- COL information is described in DCD Chapter 19.3.

- Connection points for FLEX equipment are incorporated in the system figures along with Table 6-1, which identifies the external connection components.

- Installed safety related pumps and valves are summarized in Table 6-2, with cross references to the DCD.

- On-Site and Off-Site FLEX equipment are summarized in Table 6-3 and Table 6-4, respectively with cross references to the DCD.

6.2 Specific Design Enhancements and Programs 6.2.1 Beyond Design Basis Seismic and Flood Protection BDB seismic and flood protection is a COL item.

6.2.2 Primary Side FLEX Pump(s) and Connections 6.2.2.1 Design Description One primary side FLEX pump connection has been provided into the SIS, downstream of the safety injection pump (SIP) no. 1 discharge line connection to the direct vessel injection (DVI) nozzle on the reactor vessel (RV) in the RCS, as shown in Figure 6-1. The primary side FLEX pump connection can be used by the high-head or low-head FLEX pump, depending on their necessity. The primary side high-head FLEX pump suction is the IRWST, while the low-head FLEX pump suction will be determined by the COL applicant depending on the site specific FLEX strategy for low mode operations with SGs not available. The connector size to the hose screw connector upstream of the primary FLEX pump suction is designed as 6.35 cm(2.5 in) diameter in accordance with the fire industry standard, while the primary FLEX pump suction line is designed as 10.16 cm (4 in) diameter. The connection for FLEX pump will not introduce new failure during normal plant operation by keeping the RCS pressure boundary through manual isolation (Safety Class 1) and blind flange.

KEPCO & KHNP 100

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 6.2.2.2 Design Basis The IRWST is used as the water source for the ACP, and the primary side high-head FLEX pump. The water volume required for RCS inventory makeup during Phase 2 of full-power operation is approximately 3

643.52 m (170,000 gal). The onsite water sources are sufficient to maintaining the plant in hot standby or hot shutdown condition for 2 weeks without considering consumption for the SFP cooling.

The primary side high-head FLEX pump is designed to supply 189.25 L/min (50 gpm) constantly, regardless of RCS pressure, in order to maintain the RCS inventory and remain in the hot shutdown condition, if the event occurs during full-power operation or lower mode of operation with SGs available.

2 Alternatively, the low-head FLEX pump is designed to have a TDH of 160.02 m (525 ft) (17 kg/cm A [243 psia] approximately) at 2,839 L/min (750 gpm) in order to maintain the RCS inventory and keep the cold shutdown condition by feed-and-bleed at lower modes of operation with SGs not available.

The primary side FLEX pump will be designed, manufactured, tested and installed in accordance with the applicable commercial codes and standards such as hydraulic institute pump standard, and with the design, storage, maintenance, testing as outlined in the NEI 12-06, Rev 0 Section 11.0. The primary side FLEX pump will not be designed as Seismic Category I, but will be stored in structures which are designed to satisfy GDC 2 to ensure meeting functional requirements for external environments such as seismic, flooding, wind, etc. for the specific site.

The installed non-safety related in-line valves for the safety injection system are designed as Quality Group D and Seismic Category I requirements. Since these are in-line valves, there is no specific regulating performance requirement for these components. The applicable codes and standards for valves are ASME B31.1 and B16.34 The COL applicant is responsible for determining the final FLEX pump design head considering site conditions.

6.2.2.3 Compliance with NRC Recommendation By incorporating this design into the APR1400, an alternate strategy of providing RCS inventory makeup is available when the ACP is not available. This core cooling strategy is described as the contingency plan in Subsection 5.1.2.3 of this report. This design change increases the reliability of the IRWST to maintain RCS water inventory after a BDBEE. This design feature complies with the requirements specified in References 5, 7, and 8.

6.2.3 Spent Fuel Pool - Makeup Line and Spray Line Enhancements 6.2.3.1 Design Description As part of the FLEX strategy to address Recommendation 4.2, Figures 6-2, 6-3, and 6-4 depict the SFP configuration to maintain SFP cooling by providing SFP makeup and SFP spray capabilities. Therefore, the following design is provided in the APR1400 to enhance the capability of the SFP diverse makeup lines and SFP spray lines to cope with BDBEEs:

Primary Connection: Permanently installed suction connection from the RWT for FLEX pump suction.

RWT is used as the suction water source of the FLEX pumps. Two seismically qualified, 15.24 cm(6 in) diameter lines are installed downstream of RWT in the yard. The primary and secondary piping connections with isolation valves are located outside building and hose KEPCO & KHNP 101

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 connector are located in the yard. A 15.24 cm(6 in) flexible hose with proper fittings is connected between the water supply line and the FLEX pump suction.

Hose connections are provided for the FLEX pump connections for the SFP spray lines and SFP diverse makeup standpipes at the exterior of the auxiliary building.

6.2.3.2 Design Basis The SFP spray pump and SFP makeup pump will be designed, manufactured, tested and installed in accordance with the applicable commercial codes and standards such as hydraulic institute pump standard and with the design, storage, maintenance, testing as outlined in the NEI 12-06, Rev 0 Section 11.0. The SFP spray pump and SFP makeup pump will not be designed as Seismic Category I, but will be stored in structures which are designed to satisfy GDC 2 to ensure meeting functional requirements for external environments such as seismic, flooding, wind, etc. for the specific site.

The installed non-safety related in-line valves for the SFP external makeup water system are designed as Quality Group D, and Seismic Category I requirements. Since these are in-line valves, there is no specific regulating performance requirement for these components.The applicable codes and standards for valves are ASME B31.1 and B16.34.

The SFP diverse makeup and spray lines are 15.24 cm (6 in) and 10.16 cm (4 in) pipes, respectively, to accommodate the makeup flow of 1,893 L/min (500 gpm) at a discharge pressure head of 32 m (105 ft),

and the spray flow of 757 L/min (200 gpm) at a discharge pressure head of 32.6 m (107 ft). Since a flow rate of 493.2 L/min (130.3 gpm), approximately, is required to restore SFP inventory during SFP boiling (see Subsection 5.1.2.4), pipe sizes for the SFP makeup and spray lines are sufficient to provide the necessary flow rate during BDBEE.

These seismically qualified SFP makeup and SFP spray lines are connected to an onsite source of water, namely, the RWT.

The COL applicant is responsible for determining the final FLEX pump design head considering site conditions.

6.2.3.3 Compliance with NRC Recommendations By incorporating this design into the APR1400, diverse and reliable sources of makeup water to the SFP are available to cope with BDBEE. This SFP cooling strategy is described in Subsection 5.1.2.4 of this report. These design features comply with the requirements specified in References 5, 7, and 8.

6.2.4 SFP Level Instrumentation 6.2.4.1 Design Description The key SFP water levels associated with this design are described in Subsection 5.1.3. The specific design description is as follows:

Level 1: Level adequate to support operation of the normal SFP cooling system Indicated water level on either the primary or backup instrument channel of greater than El. 144 ft 0 in (based on ensuring the open end of the normal suction lines does not become uncovered) plus the accuracy of the SFP water level instrument channel.

Level 2: Level adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck KEPCO & KHNP 102

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 The indicated level on either the primary or backup instrument channel of greater than 3.05 m (+/-0.305 m)

(10 ft [+/-1 ft]) above the top of the fuel storage racks. The 3.05 m (10 ft) criterion is conservative with regard to dose, in that the APR1400 DCD Subsections 9.1.3.1 and 9.1.3.3.4 indicate that dose would remain at or below 0.025 mSv (2.5 mrem/hr) at the surface of the water. This monitoring level provides reasonable assurance there is adequate water level to provide substantial radiation shielding for a person standing on the SFP operating deck. The elevation associated with this level is greater than 139 ft 8 in plus the accuracy of the SFP water level instrument channel, which will be determined at the COL stage.

Level 3: Level where fuel remains covered and actions to implement makeup water addition should no longer be deferred The indicated level is on either the primary or backup instrument channel of greater than 0.305 m (1 ft) above the top of the fuel storage racks. The elevation associated with this level is greater than 129 ft 8 in plus the accuracy of the SFP water level instrument channel, which will be determined at the COL stage.

This monitoring level provides assurance that there is adequate water level above the stored fuel seated in the rack.

The following instruments are provided at the SFP to address the requirements of NRC JLD-ISG-2012-03 Rev. 0 and NEI 12-02 Rev. 1. Specifically, the channels are designed as described below:

Primary (fixed) Instrument Channel (Channel A)

The primary instrument channel provides level indication through the use of guided wave radar (GWR) technology using the principle of TDR. The instrument provides a single continuous span from above Level 1 to within 0.305 m (1 ft) of the top of the spent fuel racks.

Backup Instrument Channel (Channel B)

The backup instrument channel is identical to the primary channel and is a permanent, fixed channel. The backup instrument channel provides level indication through the use of GWR technology using the principle of TDR. The instrument provides a single continuous span from above Level 1 to within 0.305 m (1 ft) of the top of the spent fuel racks.

The primary and backup instrument channels provide continuous level indication over a minimum range from the high SFP alarm El. 154 ft 2 in plus the accuracy of the SFP water level instrument channel to the top of the spent fuel racks at El 129 ft 8 in minus the accuracy of the SFP water level instrument channel.

6.2.4.2 Design Basis The SFP instruments selected are seismically mounted. The probe is designed to operate in borated water and non-borated water over the entire expected range of pool conditions from normal water temperatures to boiling temperatures. Cables and connections are designed for expected radiation levels and environments of greater than 100 °C (212 °F) and 100 percent humidity.

6.2.4.3 Compliance with NRC Recommendations The requirements and guidelines of NEI 12-02, Rev. 1 and NRCs JLD-ISG-2012-03, Rev. 0 are met.

6.2.5 AFWS Secondary Side FLEX Pump Connection 6.2.5.1 Design Description Two secondary side diesel-driven FLEX pump connections are provided to the auxiliary feedwater system (AFWS) supply lines. One FLEX pump is connected to the train of the TDAFWP PP01A and the other to KEPCO & KHNP 103

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 the TDAFWP PP001B. The FLEX pump suction and discharge pipes are 15.24 cm (6 in) diameter with Siamese connection. The suction and discharge connections are provided at the upstream of the auxiliary feedwater pump (AFWP) suction and the upstream of the AFW modulating valve, respectively. The RWT is an alternate water source that is independent, seismically qualified, and is connected to the AFWP suction. The piping sections connected at the AFW supply lines are classified as Safety Class 3, seismic Category I. The piping section downstream of the isolation valve at the exterior of the auxiliary building up to the connector is non-safety class and designed as seismic Category I. The specific features are depicted in Figure 6-5. Also, Figure 6-6 depicts the fuel oil connection for the secondary side FLEX pumps.

6.2.5.2 Design Basis The AFWSTs are used as the water source for the TDAFWP and the secondary side FLEX pumps. Before water in the AFWST depletes, the suction of the TDAFWP is switched to the RWT. The onsite water sources are sufficient to keep the hot shutdown condition and for continuous NCC operation for at least 11 days.

Each secondary side FLEX pump is designed to remove decay heat and keep the hot shutdown condition.

The fuel for secondary side FLEX pump is supplied from the EDG fuel oil storage tank A/B as shown in Figure 6-6.

The secondary side FLEX pump will be designed, manufactured, tested and installed in accordance with the applicable commercial codes and standards such as hydraulic institute pump standard, and with the design, storage, maintenance, testing as outlined in the NEI 12-06, Rev 0 Section 11.0. The secondary side FLEX pump will not be designed as Seismic Category I, but will be stored in structures which are designed to satisfy GDC 2 to ensure meeting functional requirements for external environment such as seismic, flooding, wind, etc. for the specific site. Each secondary side FLEX pump is designed for 1,174 L/min (310 gpm) at a TDH of 160 m (525 ft).

The COL applicants are responsible to determine the final FLEX pump design head considering site conditions.

6.2.5.3 Compliance with NRC Recommendations By incorporating this design into the APR1400, alternative water makeup sources to the TDAFWP are available to supplement the water source to the SG and provide RCS cooldown. This design increases the reliability of the AFWS to maintain the reactor core cooldown after the BDBEE. This design feature and the mitigation strategies addressed herein comply with the requirement specified in References 5, 7, and 8.

6.2.6 Electric Power Supply System 6.2.6.1 Design Description - Electrical Systems The APR1400 has one 4.16 kVac, 5,000 kW and two 480 Vac, 1,000 kW mobile gas turbine generators for the N+1 requirement, and those mobile GTGs are designed to meet the load requirements.

Subsection 5.1.2.6 provides the load list for each mobile GTG required during a BDBEE. The 480 V mobile GTG is credited to power the Class 1E 480 V load center during Phase 2, while the 4.16 kV mobile GTG is credited to power the Class 1E switchgear during Phase 3.

During Phase 2, the onsite EDG fuel oil storage tanks are used as the source of fuel for the mobile GTGs.

The capacity of the each EDG fuel oil storage tank is designed to allow the mobile GTG to operate at rated power for 7 days. During Phase 3, the fuel oil source is provided by offsite sources.

KEPCO & KHNP 104

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 The 4.16 kV mobile GTG is connected to the 4.16 kV switchgear Train A (or B), and the 480 V mobile GTG is connected to 480 V load center Train A (or B). The provisions to connect these GTGs are incorporated in the APR1400 design. The 4.16 kV GTG powers the 4.16 kV switchgear, 480 V load center and motor control center, 480 Vac / 125 Vdc battery charger, 125 Vdc battery, 125 Vdc / 120 Vac inverter, and 120 Vac distribution panel in Train A (or B). The 480 V mobile GTG powers the 480 V load center and motor control center, 480 Vac / 125 Vdc battery charger, 125 Vdc battery, 125 Vdc/120 Vac inverter, and 120 Vac distribution panel in Train A (or B).

During an ELAP, Class 1E 125 Vdc power is required for operation of 4.16 kV switchgears, 125 Vdc loads, 480 Vac MOVs and AOVs that are 125 Vdc battery backed up, I&C panels and shutdown system instrumentation, and 120 Vac loads that are inverted from 125 Vdc batteries.

Emergency Lighting and Communication Emergency lighting in the areas such as MCR and TSC/OSC is provided from the plant Class 1E batteries during Phase 1 and mobile GTG during Phase 2. Electric power to the communication subsystem is provided from the 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months dedicated emergency UPS and ELAP GTG during Phases 1 and 2, respectively. Accordingly, the APR1400 design addresses the emergency lighting and communication during BDBEE.

6.2.6.2 Design Basis The following mobile generators are used as the power source of Train A or train B power system during BDBEE:

In Phase 2, two onsite 480V mobile GTGs are provided to meet N+1 requirement and connected to either train of the 480V load center. Each 480V mobile GTG has a capacity of 1,000 kW.

In Phase 3, one offsite 4.16 kV mobile GTG is provided and connected to either train of the Class 1E 4.16 kV switchgear. The 4.16 kV mobile GTG has a capacity of 5,000 kW.

6.2.7 Operational Program, Procedures, and Training The programs, procedures, guidance, and training addressing EOP/SAMGs/EDMGs for BDBEE are a COL item.

6.2.8 Emergency Procedures The emergency communication system/enhancement, staffing large-scale natural events, and revisions to EP for ELAP are COL items.

6.2.9 Storage of FLEX Equipment The COL applicant is responsible for addressing the details of the following guidance for the storage and deployment of the FLEX equipment:

1. The FLEX equipment is stored in dedicated buildings/structures that will provide reasonable accessibility and withstand the BDBEEs and meet the requirements of 10 CFR 50, Appendix A, GDC 2. The N+1 equipment is stored in separate buildings.

2. Suitable storage locations that provide reasonable protection during specific external events are selected in accordance with guidance provided in NEI 12-06 (Reference 8) Section 5 through 9.

3. A technical basis should be established for the storage for the FLEX equipment that provides the KEPCO & KHNP 105

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 inputs, assumptions, and documented basis that the mitigation strategy and support equipment will be reasonably protected from applicable external events such that the equipment could be operated in place, if applicable, or moved to its deployment locations. This basis should be auditable, consistent with generally accepted engineering principles, and controlled within the configuration document control system.

4. The FLEX equipment should be stored in storage locations chosen for the equipment that provide protection from external events as necessary to allow the equipment to perform its function without loss of capability such that no one external event can reasonably fail the site FLEX capability.

5. Storage locations must include considerations of a suitable and convenient means to bring the equipment to the connection points in time to initiate the strategy prior to expiration of the estimated capabilities to maintain core and spent fuel pool cooling, and containment functions in the initial response phase. Consideration should be given to the transport from the storage areas following the external event, recognizing that external events can result in obstacles restricting normal pathways for movement.

6. If the FLEX equipment is pre-staged such that it minimizes the time delay and burden of hook-up following an external event, then the equipment should be evaluated to not have an adverse effect on existing SSCs and the primary connection point should be as close to the intended point of supply as possible.

7. The FLEX equipment should be stored and maintained in a manner that is consistent with assuring that it does not degrade over long periods of storage and that it is accessible for periodic maintenance and testing.

8. If 50.54(hh)(2) equipment is credited in the FLEX mitigating strategies, it should meet the above storage requirements in addition to the 50.54(hh)(2) requirements.

9. Storage of debris removal equipment (if needed) should provide reasonably protection from the applicable external events such that it remains functional and deployable to clear obstructions from the pathway between the FLEX equipments storage location and its deployment location(s).

10. Deployment of the FLEX equipment or debris removal equipment from storage locations should not depend on offsite power or onsite emergency ac power (e.g., to operate roll up doors, lifts, elevators, etc.).

6.2.10 Installed equipment and tanks utilized in the mitigation strategies All permanent installed, safety related equipment (pumps, valves, etc.) that is utilized in the mitigation strategies for BDBEE are housed inside the reactor containment building, auxiliary building, essential service water/component cooling water heat exchanger building, emergency diesel generator building. All of these structures are safety related and are designed for seismic, flood, high wind and missile. The specific location, function, and classification of these structures are described in DCD Tier 2, Table 3.2-1.

The specific tanks (AFWST, IRWST, SIT, EDG fuel oil storage tank, EDG fuel oil day tank, BAST) are utilized in the mitigation strategies. All of these tanks are safety related, seismic Category I, and Quality Group C. DCD Tier 2, Table 3.2-1 provides the location, function, and safety classifications for these tanks.

Additionally, RWT is utilized for mitigating strategies and is designed to seismic Category I and Quality Group D. It will remain functional for the mitigating strategies. The COL applicant will confirm and ensure that the RWT and flow path to the FLEX equipment(structures, piping, components, and connections) are designed to be robust with respect to applicable hazards (e.g., seismic events, floods, high winds, and missiles).

KEPCO & KHNP 106

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 6.2.11 Connections for FLEX strategies The detailed design of connections for FLEX strategies located outside the buildings is the responsibility of COL applicant. The COL applicant is to ensure that all the connections (See Table 6-5) for FLEX strategies located outside the buildings are accessible and protected from all applicable external hazards (e.g., seismic events, floods, high winds, and associated missiles).

KEPCO & KHNP 107

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 6-1 External Connection Components for BDBEE (1 of 2)

Component DCD Chapter and/or Section Function V2601 Figure 9.1.3-1 SFP external makeup line check valve V2602 Figure 9.1.3-1 SFP external makeup line isolation valve V2605 Figure 9.1.3-1 SFP external spray line check valve V2606 Figure 9.1.3-1 SFP external spray line isolation valve V2611 Figure 9.1.3-1 SFP external makeup line check valve V2612 Figure 9.1.3-1 SFP external makeup line isolation valve V2615 Figure 9.1.3-1 SFP external spray line check valve V2616 Figure 9.1.3-1 SFP external spray line isolation valve SI-801 Table 3.9-4, Table 3.9-13, External emergency injection line check valve Figure 6.3.2-1 (4 of 4)

SI-803 Table 3.9-4, Table 3.9-13, External emergency injection line isolation valve Figure 6.3.2-1 (4 of 4)

SI-805 Figure 6.3.2-1 (4 of 4) External emergency injection line fill isolation valve SI-807 Figure 6.3.2-1 (4 of 4) External emergency injection line isolation valve CH-784 Figure 9.3.4-1 (4 of 7) Primary side high-head FLEX pump suction isolation V2678A Figure 10.4.9-1 AF FLEX pump suction line backflow prevention V2678B Figure 10.4.9-1 AF FLEX pump suction line backflow prevention V2679A Figure 10.4.9-1 AF FLEX pump suction line isolation V2679B Figure 10.4.9-1 AF FLEX pump suction line isolation V2098A Figure 10.4.9-1 AF FLEX pump discharge line backflow prevention V2098B Figure 10.4.9-1 AF FLEX pump discharge line backflow prevention V2102A Figure 10.4.9-1 AF FLEX pump discharge line isolation V2102B Figure 10.4.9-1 AF FLEX pump discharge line isolation V2001A Figure 9.5.4-1 Diesel fuel oil day tank discharge line to mobile equipment header isolation V2001B Figure 9.5.4-1 Diesel fuel oil day tank discharge line to mobile equipment header isolation V2202A Figure 9.5.4-1 Diesel fuel oil supply line to mobile GTG isolation V2202B Figure 9.5.4-1 Diesel fuel oil supply line to mobile GTG isolation V2203A Figure 9.5.4-1 Diesel fuel oil supply line to AF FLEX pump isolation V2203B Figure 9.5.4-1 Diesel fuel oil supply line to AF FLEX pump isolation V2204A Figure 9.5.4-1 Diesel fuel oil supply line to primary high-head pump isolation V2204B Figure 9.5.4-1 Diesel fuel oil supply line to primary high-head pump isolation KEPCO & KHNP 108

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 6-1 External Connection Components for BDBEE (2 of 2)

Component DCD Chapter and/or Section Function V2205A Figure 9.5.4-1 Diesel fuel oil supply line to primary low-head pump isolation V2205B Figure 9.5.4-1 Diesel fuel oil supply line to primary low-head pump isolation V2206A Figure 9.5.4-1 Diesel fuel oil supply line to SFP pump isolation V2206B Figure 9.5.4-1 Diesel fuel oil supply line to SFP pump isolation V2207A Figure 9.5.4-1 Diesel fuel oil supply line to SFP spray pump isolation V2207B Figure 9.5.4-1 Diesel fuel oil supply line to SFP spray pump isolation V2208A Figure 9.5.4-1 Diesel fuel oil supply line to ECSBS FLEX pump V2208B Figure 9.5.4-1 Diesel fuel oil supply line to ECSBS FLEX pump Circuit Breaker of Figure 8.1-1 (1 of 2) Provision for connecting to 4.16 kV mobile generator Class 1E 4.16 kV Switchgear 01A (1-823-E-SW01A)

Circuit Breaker of Figure 8.1-1 (2 of 2) Provision for connecting to 4.16 kV mobile generator Class 1E 4.16 kV Switchgear 01B (1-823-E-SW01B)

Circuit Breaker of Figure 8.1-1 (1 of 2) Provision for connecting to 480V mobile generator Class 1E 480 V Load Center 01A (1-825-E-LC01A)

Circuit Breaker of Figure 8.1-1 (2 of 2) Provision for connecting to 480V mobile generator Class 1E 480 V Load Center 01B (1-825-E-LC01B)

Battery 9.5.2.1 The communication systems are powered from one of the two dedicated 16-hour-rated non-safety-related batteries (normal and standby) in case of either AAC GTG failure during a LOOP or SBO condition.

KEPCO & KHNP 109

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 6-2 List of Installed Safety Related Pumps and Valves for BDBEE (1 of 2)

Description DCD Tier 2 Reference Turbine driven Auxiliary Feedwater Pump 10.4.9, Table 10.4.9-1 Main Steam Safety Valve 10.3.2, Table 10.3.2-1 Main Steam Atmospheric Dump Valve 10.3.2, Table 10.3.2-1 Auxiliary Charging Pump 9.3.4, Table 9.3.4-2 Boric Acid Makeup Pump 9.3.4, Table 9.3.4-2 Boric Acid Makeup Pump 9.3.4, Table 9.3.4-2 Direct Boration Valve 9.3.4,

BAST Gravity Valve (Train A) 9.3.4,

BAST Gravity Valve (Train B) 9.3.4,

Charging Containment Isolation Valve 9.3.4, Table 6.2.4-1 Diesel Fuel Oil Transfer Pump 1-595-PP01A 9.5.4, Table 9.5.4-1 Diesel Fuel Oil Transfer Pump 1-595-PP02A 9.5.4, Table 9.5.4-1 Diesel Fuel Oil Transfer Pump 1-595-PP01B 9.5.4, Table 9.5.4-1 Diesel Fuel Oil Transfer Pump 1-595-PP02B 9.5.4, Table 9.5.4-1 Shutdown Cooling Pump-1 5.4.7, Table 5.4.7-1 Shutdown Cooling Pump-2 5.4.7, Table 5.4.7-1 SCS Suction Line Isolation Valve SI-651 5.4.7,

SCS Suction Line Isolation Valve SI-652 5.4.7,

SCS Suction Line CTMT Isolation Valve SI-655 5.4.7, Table 6.2.4-1 SCS Suction Line Isolation Valve SI-656 5.4.7,

SCS Warmup Line Isolation Valve SI-690 5.4.7,

SCS Warmup Line Isolation Valve SI-691 5.4.7,

SCS Test Return Line Isolation Valve SI-314 5.4.7,

SCS Test Return Line Isolation Valve SI-315 5.4.7,

SCS Test Return Line Isolation Valve SI-688 5.4.7,

SCS Test Return Line Isolation Valve SI-693 5.4.7,

SDCHX Bypass Flow Control Valve SI-312 5.4.7,

SDCHX Bypass Flow Control Valve SI-313 5.4.7,

SDCHX Discharge Isolation and Throttle Valve SI-310 5.4.7,

SDCHX Discharge Isolation and Throttle Valve SI-311 5.4.7,

IRWST Return Line CTMT Isolation Valve SI-300 5.4.7, Table 6.2.4-1 IRWST Return Line CTMT Isolation Valve SI-301 5.4.7, Table 6.2.4-1 SCS DVI CTMT Isolation Valve SI-600 5.4.7, Table 6.2.4-1 SCS DVI CTMT Isolation Valve SI-601 5.4.7, Table 6.2.4-1 IRWST Return Line Isolation Valve SI-395 5.4.7,

SIP/SCP Suction Cross Connection Valve SI-344 5.4.7,

SIP/SCP Suction Cross Connection Valve SI-346 5.4.7,

CCW Pump 1A 9.2.2, Table 9.2.2-4 CCW Pump 1B 9.2.2, Table 9.2.2-4 CCW HX HE01A Outlet Valve 1-461-V-0021 9.2.2, Table 9.2.2-5 CCW HX HE01B Outlet Valve 1-461-V-0022 9.2.2, Table 9.2.2-5 CCW HX HE02A Outlet Valve 1-461-V-0023 9.2.2, Table 9.2.2-5 KEPCO & KHNP 110

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 6-2 List of Installed Safety Related Pumps and Valves for BDBEE (2 of 2)

Description DCD Tier 2 Reference CCW HX HE02B Outlet Valve 1-461-V-0024 9.2.2, Table 9.2.2-5 CCW HX HE03A Outlet Valve 1-461-V-0025 9.2.2, Table 9.2.2-5 CCW HX HE03B Outlet Valve 1-461-V-0026 9.2.2, Table 9.2.2-5 CCW HX Bypass ISOL Valve 1-461-V-0027 9.2.2, Table 9.2.2-5 CCW HX Bypass ISOL Valve 1-461-V-0028 9.2.2, Table 9.2.2-5 Non-Safety Load ISOL Valve 1-461-V-0143 9.2.2, Table 9.2.2-5 Non-Safety Load ISOL Valve 1-461-V-0144 9.2.2, Table 9.2.2-5 Non-Safety Load ISOL Valve 1-461-V-0149 9.2.2, Table 9.2.2-5 Non-Safety Load ISOL Valve 1-461-V-0150 9.2.2, Table 9.2.2-5 DG A HX Inlet ISOL Valve 1-461-V-0191 9.2.2, Table 9.2.2-5 DG B HX Inlet ISOL Valve 1-461-V-0192 9.2.2, Table 9.2.2-5 RCP Cooling Return CTMT ISOL Valve 1-461-V-0249 9.2.2, Table 6.2.4-1 Letdown HX Cooling Supply CTMT ISOL Valve 1-461-V-0296 9.2.2, Table 6.2.4-1 Letdown HX Cooling Supply CTMT ISOL Valve 1-461-V-0297 9.2.2, Table 6.2.4-1 Letdown HX Cooling Return CTMT ISOL Valve 1-461-V-0301 9.2.2, Table 6.2.4-1 Letdown HX Cooling Return CTMT ISOL Valve 1-461-V-0302 9.2.2, Table 6.2.4-1 SC HX 01A Inlet ISOL Valve 1-461-V-0351 9.2.2,

SC HX 01B Inlet ISOL Valve 1-461-V-0352 9.2.2,

ESSEN CHLR CNDSR 1A Outlet ISOL Valve 1-461-V-0383 9.2.2,

ESSEN CHLR CNDSR 1B Outlet ISOL Valve 1-461-V-0384 9.2.2,

SFP Cooling HX 02A Inlet ISOL Valve 1-461-V-0389 9.2.2,

SFP Cooling HX 02B Inlet ISOL Valve 1-461-V-0390 9.2.2,

Essential Service Water (ESW) Pump 1A 9.2.1, Table 9.2.1-1 Essential Service Water (ESW) Pump 1B 9.2.1, Table 9.2.1-1 ESW PP 01A DISCH Valve 1-462-V-0045 9.2.1,

ESW PP 01B DISCH Valve 1-462-V-0046 9.2.1,

ESW Flow Control Valve 1-462-V-0071 9.2.1,

ESW Flow Control Valve 1-462-V-0072 9.2.1,

Cooling Tower Bypass Valve 1-462-V-0075 9.2.1,

Cooling Tower Bypass Valve 1-462-V-0076 9.2.1,

MSADV Inlet Isolation Valve 1-521-V105 10.3.2,

MSADV Inlet Isolation Valve 1-521-V106 10.3.2,

MSADV Inlet Isolation Valve 1-521-V107 10.3.2,

MSADV Inlet Isolation Valve 1-521-V108 10.3.2,

Aux. Feedwater CTMT Isolation Valve 1-542-V-0043 10.4.9, Table 6.2.4-1 Aux. Feedwater CTMT Isolation Valve 1-542-V-0044 10.4.9, Table 6.2.4-1 Essential Chilled Water Pump 1A 9.2.7, Table 9.2.7-1 Essential Chilled Water Pump 1A 9.2.7, Table 9.2.7-1

These valves are inline valves, and there is no unique performance requirement for these valves.

KEPCO & KHNP 111

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 6-3 List of On-site FLEX Equipment for BDBEE (1 of 2)

Item Interface Design Functional Description Quantity (Note) Reference No parameters Requirements 1 Primary side Two (2)

Diesel driven Supply makeup DCD Tier 2 high-head FLEX

Flowrate: water to RCS Subsection pump 189.25 L/min (50 gpm) when ACP is not 19.3.2.3.1.1

Operating Pressure: available.

2 105.46 kg/cm A TeR (1,500 psia) Sections

Water Source: 5.1.2.3.1.2.2 IRWST & 6.2.2.2 2 Primary side Two (2)

Diesel driven Supply makeup DCD Tier 2 low-head FLEX

Flowrate: water to RCS Subsection pump 2,839 L/min(750 gpm) during Phase 2 19.3.2.3.1.3

TDH: when SGs are not 160 m (525 ft) available. TeR

Water Source: Sections Available borated water 5.1.2.3.3.2 source (COL 19.3(17)) & 6.2.2.2 3 Secondary side Three (3)

Diesel driven Supply cooling DCD Tier 2 FLEX pump

Flowrate: water to SGs Subsection 1,174 L/min (310 gpm) when TDAFWP is 19.3.2.3.1.1

TDH: not available.

160 m (525 ft) TeR

Water Source: Sections AFWST and RWT 5.1.2.3.1.2.2

& 6.2.5.2 4 SFP Makeup One (1)

Diesel driven Supply makeup DCD Tier 2 FLEX pump

Flowrate: water to SFP. Subsection 1,893 L/min (500 gpm) 19.3.2.3.2

Discharge Pressure:

32 m (105 ft) TeR

Water Source: Sections RWT 5.1.2.4.1.2

& 6.2.3.2 5 SFP Spray One (1)

Diesel driven Supply makeup DCD Tier 2 FLEX pump

Flowrate: water to SFP. Subsection 757 L/min (200 gpm) 19.3.2.3.2

Discharge Pressure:

32.6 m (107 ft) TeR

Water Source: Sections RWT 5.1.2.4.1.2

& 6.2.3.2 6 ECSBS One (1)

Diesel driven Supply water to DCD Tier 2 FLEX Pump

Flowrate: containment Subsection 2,839 L/min (750 gpm) atmosphere to 19.3.2.3.3

Discharge Pressure: prevent 200 m (656 ft) containment TeR

Water Source: overpressurization Sections RWT during low mode 5.1.2.5 operation.

KEPCO & KHNP 112

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 6-3 List of On-site FLEX Equipment for BDBEE (2 of 2)

Item Interface Design Functional Description Quantity (Note) Reference No parameters Requirements 7 480 V Two (2) 1000 kW each Supply power to DCD Tier 2 mobile GTG 480 V load center, Subsection motor control 19.3.2.3.1.1 center, and 125 Vdc battery TeR charger via 480 V Sections Class 1E power 5.1.2.6.1.1, system Train A or B 6.2.6, and during Phase 2. Table 5-5 (Note) The COL applicant is responsible for determining the final FLEX pump design head considering site conditions.

Table 6-4 List of Off-site FLEX Equipment for BDBEE Item Interface Design Functional Description Quantity Reference No parameters Requirements 1 4.16 kV One (1) 5000 kW Supply power to 4.16 DCD Tier 2 mobile GTG kV switchgear and Subsection etc., and restore Train 19.3.2.3.1.1 A or B of the Class 1E power system during TeR Phase 3. Sections 5.1.2.6.1.1, 6.2.6, and Table 5-5 KEPCO & KHNP 113

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table 6-5 List of Connection for FLEX strategies Flex Equipment Connections Quantity Location Functional Requirement Primary side high- Discharge 1 Outside To supply makeup water to RCS head or primary side connection Building low-head FLEX pump SFP makeup FLEX Discharge 2 Outside To supply makeup water to SFP pump connections Building SFP spray FLEX Discharge 2 Outside To supply makeup water to SFP pump connections Building ECSBS FLEX pump Discharge 1 Outside To prevent containment over-connection Building pressurization Secondary side FLEX Discharge 2 Outside To supply cooling water to pump connections Building associated SG Secondary side FLEX Suction 2 Outside To supply cooling water to pump connections Building associated SG Primary side high- Suction 1 Outside To supply makeup water to RCS head FLEX pump connection Building FLEX pumps and Fuel oil 14 Outside To supply fuel oil to FLEX pumps mobile GTGs supplying Building and mobile GTG connections The 480 V mobile Electrical 2 COL To supply the power to Class 1E GTG connections 19.3(10) 480 V load centers The 4.16 kV mobile Electrical 2 COL To supply the power to 4.16 kV GTG connections 19.3(10) switchgear KEPCO & KHNP 114

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Figure 6-1 Primary Side FLEX Pump Connection into the Safety Injection System KEPCO & KHNP 115

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Figure 6-2 FLEX Pump Suction Source for SFP Makeup and Spray Line AB RCB N

FUEL TRANSFER CANAL V1015 V1000 FUEL TRANSFER TUBE EL. 114'-6 052A-4 V1101 006AA-4 CASK 004B-4 E LOADING T A

PIT G EL. 111'-0 004AD-2 GATE 4 INCH SPRAY LINE 004AE-2 001AC-10 10"X16" 16"X10" 6 INCH 001AD-10 MAKEUP LINE 10"X16" 16"X10" (NOTE 2)

AB YARD SPENT FUEL POOL EL. 114'-0 DEMIN. WATER 004AF-2 001AA-10 1582-005 004AC-2 (NOTE 2) H9 003AA-14 4 INCH SPRAY LINE 004AB-2 001AB-10 6 INCH MAKEUP LINE V1105 004AA-2 6"X4" 006AB-4 YARD AB 4"X6" 001BB-16 V1018 056A-3 004B-4 V1106 Figure 6-3 Connection for SFP Makeup and Spray Line KEPCO & KHNP 116

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Figure 6-4 Layout of SFP Makeup and SFP Spray Line Connections KEPCO & KHNP 117

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Figure 6-5 Water Supply System to the Secondary Side of SG Figure 6-6 Fuel Oil Supply System to FLEX Pumps KEPCO & KHNP 118

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 AUXILIARY BUILDING Raw Water Tank 4 INCH DISCHARGE 4X6 6"X4" 10 INCH SUPPLY CS-V1014 CS-V1013 6 INCH 4 INCH LINE 4X6 4 INCH 6 INCH WATER SOURCE LINE CONTAINMENT FLEX PUMP FROM RWT BUILDING Figure 6-7 Flow Path for FLEX Connection to Deliver Water to Containment for ECSBS KEPCO & KHNP 119

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1

7.0 CONCLUSION

The APR1400 strategy and design enhancement along with regulatory recommendations and regulatory requirements related to lessons learned from the accident at Fukushima Dai-ichi Nuclear Power Plant after the 2011 Great Tohoku Earthquake on March 11, 2011 are described in this Technical Report. The impact assessments to the current APR1400 DCD are prepared. Multiple design enhancements are identified and have been implemented with the objective to enhance mitigation capability to beyond-design-basis external events. The design enhancements described in this report increase the APR1400 reliability for safety against the BDBEE. The operational and programmatic aspects of responding to the BDBEE CCW will be addressed by COL applicants prior to the initial fuel loading.

KEPCO & KHNP 120

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1

8.0 REFERENCES

st

1. SECY-11-0093, Recommendations for Enhancing Reactor Safety in the 21 Century, U.S.

Nuclear Regulatory Commission, July 2011.

2. SECY-11-0137, Prioritization of Recommended Actions to be Taken in Response to Fukushima Lessons Learned," U.S. Nuclear Regulatory Commission, October 2011.

3. SECY-12-0025, "Proposed Orders and Requests for Information in Response to Lessons Learned from Japan's March 11, 2011, Great Tohoku Earthquake and Tsunami," U.S. Nuclear Regulatory Commission, February 2012.

4. SECY-12-0095, Tier 3 Program Plans and 6-Month Status Update in Response to Lessons Learned from Japans March 11, 2011, Great Tohoku Earthquake and Subsequent Tsunami, U.S. Nuclear Regulatory Commission, July 2012.

5. Order EA-12-049, "Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events," U.S. Nuclear Regulatory Commission, March 12, 2012.

6. Order EA-12-051, "Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation," U.S. Nuclear Regulatory Commission, March 2012.

7. JLD-ISG-2012-01 Compliance with Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events, Rev. 0, U.S.

Nuclear Regulatory Commission, August 29, 2012.

8. NEI 12-06, Diverse and Flexible Coping Strategies (FLEX) Implementation Guide, Rev. 0, Nuclear Energy Institute, August 2012.

9. APR1400-A-M-NR-14002-P, Extended Loss of AC Power Capability for APR1400 KSB RCP Seals, Rev.1, KHNP, January 2017.

10. NEI White Paper, Battery Life Issue, Nuclear Energy Institute, August 2013.

11. NRC Letter dated September 2013 to NEI endorsing Reference 10 with clarifications.

12. NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, Rev. 1, Nuclear Energy Institute, August 2012.

13. JLD-ISG-2012-03, Compliance with Order EA-12-051, Reliable Spent Fuel Pool Instrumentation, Rev. 0, U.S. Nuclear Regulatory Commission, August 2012.

14. IEEE Std 344-2004, IEEE Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, June 2005.

15. Enclosure 5 to "Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident," March 2012.

16. NEI 12-01, Guideline for Assessing Beyond Design Basis Accident Response Staffing and Communication Capabilities, Rev. 0, Nuclear Energy Institute, May 2012.

KEPCO & KHNP 121

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 APPENDIX. A Supporting Analysis Results for the Operational Strategy for Core Cooling KEPCO & KHNP Ai

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 TABLE OF CONTENTS A.1 INTRODUCTION ................................................................................................ A1 A.2 COMPUTER CODE ............................................................................................ A1 A.3 ACCEPTANCE CRITERIA .................................................................................. A1 A.4 ANALYSIS CONDITION AND ASSUMPTION ...................................................... A1 A.4.1 Power Operation and Shutdown Operations with SGs Available ................................................A1 A.4.2 Shutdown Operations with SGs not Available .............................................................................A2 A.5 ANALYSIS RESULTS ......................................................................................... A2 A.5.1 Full-Power Operation ...................................................................................................................A2 A.5.1.1 Core Cooling - Full-Power Operation Basic Strategy of RCS Cooldown (Using Installed Pumps) ..........................................................................................................................A2 A.5.1.2 Core Cooling - Full-Power Operation Contingency Plan (Using FLEX Pumps) .........................A3 A.5.1.3 Subcriticality - Full-Power Operation FLEX Strategy ..................................................................A4 A.5.2 Lower-Power Operation and Shutdown Conditions with SGs Available .....................................A4 A.5.2.1 Coping Capability - Mode 1 through Mode 3 ..............................................................................A4 A.5.2.2 Coping Capability - Mode 4 and Mode 5 with SGs Available .....................................................A5 A.5.3 Shutdown Condition with SGs not Available................................................................................A5 KEPCO & KHNP Aii

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 LIST OF TABLES Table A-1 Sequence of Events for Coping Operation Against ELAP/LUHS Occurred at Full-Power Operation ...........................................................................................................A6 LIST OF FIGURES Figure A-1 RCS and Steam Generator Pressure (Basic Strategy) ............................................................ 7 Figure A-2 RCS Temperature (Basic Strategy) .......................................................................................... 7 Figure A-3 SIT Flow (Basic Strategy) ......................................................................................................... 8 Figure A-4 ACP Flow (Basic Strategy) ....................................................................................................... 8 Figure A-5 RCS Leak Flow (Basic Strategy) .............................................................................................. 9 Figure A-6 RCS Pressurizer Water Level (Basic Strategy) ........................................................................ 9 Figure A-7 Collapsed Downcomer and Core Level (Basic Strategy) ....................................................... 10 Figure A-8 Cladding Temperature (Basic Strategy) .................................................................................. 10 Figure A-9 MSADV, MSSV Flow (Basic Strategy) .................................................................................... 11 Figure A-10 Integration of AFW Flow (Basic Strategy) ............................................................................. 11 Figure A-11 RCS and Steam Generator Pressure (Contingency Plan).................................................... 12 Figure A-12 RCS Temperature (Contingency Plan) ................................................................................. 12 Figure A-13 SIT Flow (Contingency Plan) ................................................................................................ 13 Figure A-14 Primary FLEX Pump Flow (Contingency Plan) ..................................................................... 13 Figure A-15 RCS Leak Flow (Contingency Plan) ..................................................................................... 14 Figure A-16 Pressurizer Water Level (Contingency Plan) ........................................................................ 14 Figure A-17 Collapsed Downcomer and Core level (Contingency Plan) ................................................. 15 Figure A-18 Cladding Temperature (Contingency Plan) ........................................................................... 15 Figure A-19 MSADV, MSSV Flow (Contingency Plan) ............................................................................. 16 Figure A-20 Integration of AFW Flow (Contingency Plan) ........................................................................ 16 Figure A-21 Reactivity Changes during RCS Cooldown (with RCP Leak) ............................................... 17 Figure A-22 Reactivity Changes during RCS Cooldown (No RCP Leak)................................................. 17 Figure A-23 RCS Pressure ....................................................................................................................... 18 Figure A-24 RCS Temperature ................................................................................................................. 18 Figure A-25 LTOP Valve Flow................................................................................................................... 19 KEPCO & KHNP Aiii

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 APPENDIX.A Supporting Analysis Results for the Operational Strategy for Core Cooling A.1 Introduction This appendix provides the supporting analyses and their results for the APR1400 core cooling capability to cope with the BDBEE, ELAP concurrent with LUHS. Specifically, the coping capability is evaluated according to the FLEX strategies described in the Subsection 5.1.2.3, for the following operation modes:

Full-power operation

Low-power operations and shutdown conditions with SGs available A.2 Computer Code A best-estimate computer code, named RELAP5/Mod 3.3, is used for the supporting analyses for the APR1400 operational strategy for core cooling. The RELAP5 code was approved by the NRC and has been widely used for the safety analysis of nuclear power plants of all over the world. The RELAP5 code is based on a non-equilibrium separated two-phase flow model and has other additional models to properly describe the thermal-hydraulic behavior of components of reactor systems including heat conduction in the core and reactor coolant system, reactor kinetics, control systems and trips.

A.3 Acceptance Criteria The following acceptance criteria based on NEI 12-06, Section 3.2.1 are applied to the supporting analysis for the operational strategy for core cooling during the ELAP concurrent with LUHS.

Core cooling is maintained

No fuel failures The fulfillment of these criteria is determined by evaluating RCS key parameters such as RCS and SG pressures, RCS temperature, collapsed levels in reactor vessel, core, and SG.

A.4 Analysis Condition and Assumption The following analysis conditions and assumptions are selected according to the requirements of NEI 12-06, Section 3.2.1.

The full-power operation is selected as a representative case for setting up the APR1400 FLEX strategy of the modes 1 through 4 and mode 5 operation with SGs available.

The plant is assumed to operate at 100 percent power with no uncertainty for system parameters.

The initiating event is assumed to be an ELAP concurrent with LUHS.

The reactor is assumed to be tripped automatically by the low reactor coolant pump speed trip of the RPS since the RCPs could not be provided with ac power.

The MSSVs are assumed to actuate automatically when the SG pressure exceeds the MSSV setpoints.

KEPCO & KHNP A1

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1

RCP seal leakage is assumed to be 94.64 L/min (25 gpm) per RCP at the initial RCS pressure of 2

158.19 kg/cm A (2,250 psia) and naturally reduce as the RCS pressure decreases during the event. This causes loss of RCS inventory, which should be adequately maintained for preventing core uncovery leading to core failure.

The TDAFWPs are assumed to start automatically on receipt of AFAS signal.

The decay heat conditions of ANSI/ANS-5.1-1979 are used for best-estimate simulation of the FLEX strategy.

Operator is assumed to cool down the RCS by controlling MSADVs with a rate of 27.78 °C/hour (50 °F/hour) from 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the BDBEE.

The auxiliary charging pump (ACP) is assumed to supply borated water at the constant value of 166.56 L/min (44 gpm) for RCS makeup after 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the event. If ACP is unavailable, a primary high-head FLEX pump is used for providing adequate water to maintain RCS inventory.

Four safety injection tanks (SITs) inject 4,000 ppm borated water into the RCS when the RCS pressure decreases below the setpoints as designed.

Normal feedwater flow to SGs is assumed to stop at the initiation of the BDBEE. Auxiliary feedwater flow supplies water to SGs and is controlled to maintain SG level within the control band of 25 percent to 40 percent A.5 Analysis Results A.5.1 Full-Power Operation The APR1400 core cooling capability under the BDBEE, ELAP concurrent with LUHS, is analyzed using RELAP5/Mod 3.3 code, according to the full-power operational strategy consisting of the following three phases as described in Subsection 5.1.2.3.1.

Phase 1: 0 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

Phase 2: 8 to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

Phase 3: Indefinite time period following Phase 2 For the full-power operation case, two types of core cooling strategies, which are basic operational strategy and contingency plan, are analyzed. In the basic strategy, the RCS is cooled down to hot shutdown using both installed plant equipment, such as TDAFWP and ACP, and FLEX equipment such as the mobile GTG. The contingency plan is prepared in case installed plant equipment is inoperable even after connection of mobile ac power. In this case, RCS is cooled down to cold shutdown using the secondary side FLEX pump. RCS makeup is carried out by the primary side high-head FLEX pump. Table A-1 summarizes the sequence of events during the ELAP/LUHS coping operation.

A.5.1.1 Core Cooling - Full-Power Operation Basic Strategy of RCS Cooldown (Using Installed Pumps)

Figures A-1 through A-9 show the primary and secondary thermal-hydraulic behaviors for the basic KEPCO & KHNP A2

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 strategy using installed equipment.

Figure A-1 shows RCS and SG pressures and Figure A-2 shows RCS temperatures during the accident.

These figures demonstrate the NCC operation of each step such as maintaining hot standby conditions, cooling down the RCS, and maintaining hot shutdown conditions. While RCS pressure decreases by secondary cooldown, the SIT starts to inject borated water when the RCS pressure reaches to the SIT actuation setpoint (Figure A-3). The ACP flow starts at 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the mobile GTG is assumed to be connected, and the flow rate is well balanced with the RCP seal leakage in the hot shutdown period (Figures A-4 and A-5).

As shown in Figure A-6, PZR level decreases and stays at an empty state for a long time. This is because the liquid volume of RCS decreases due to the RCP seal leakage and shrinkage of RCS inventory.

However, the core and downcomer water level is recovered after SIT flow injection, and then stabilized after 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months when the ACP flow becomes balanced with the RCP seal leakage.

Because of the RCS inventory makeup by SIT and ACP, the decreased core level is also restored, as shown in Figure A-7. Even though the core level reduces a little during the cooldown period, it still covers the active core. This means that the fuel is not uncovered and fuel integrity is preserved. The cladding temperature of the active core shown in Figure A-8 also shows that the fuel integrity is maintained during the accident.

Figure A-9 shows MSSV and MSADV flows during the accident. The MSSV relieves the steam generated by the decay heat of the core from SG and thus maintains the SG pressure at a constant value during the hot standby condition, whereas the MSADV is used for cooling the RCS and maintaining it at hot shutdown condition.

3 As shown in Figure A-10, the total mass of AFW flow provided during the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is about 2,513.51 m (664,000 gal), which is still below the design value of the two AFWSTs inventory.

The AFWST and the raw water tank (RWT) can be used as onsite water sources to feed the SGs, and the number of onsite water sources is sufficient for NCC operation during Phases 1 and 2.

A.5.1.2 Core Cooling - Full-Power Operation Contingency Plan (Using FLEX Pumps)

For the contingency plan using FLEX equipment, the analysis results are provided in Figures A-11 through A-20.

Figure A-11 shows RCS and SG pressures and Figure A-12 shows RCS temperatures during the accident.

These figures demonstrate the coping operation of each step of NCC operation for maintaining hot standby condition, cooling down the RCS to hot shutdown condition, further cooldown to the SG pressure 2

of 1.03 kg/cm A (14.7 psia), and maintaining the RCS at cold shutdown.

During the RCS cooldown operation, the system pressure decreases due to RCP seal leak and shrinkage.

SIT starts to inject borated water when the RCS pressure reaches the SIT actuation setpoint (Figure A-13).

The high-head primary FLEX pump makes up the RCS inventory at the flow rate of 189.27 L/min (50 gpm) from 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the event, and the makeup flow rate becomes balanced with the RCP seal leakage in the cold shutdown period (Figures A-14 and A-15). The PZR level starts to increase at around 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months due to the makeup flow from the FLEX pump, and becomes stabilized at around 17 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months when the FLEX pump flow is balanced with the RCP seal leakage, as shown in Figure A-16.

Because of the RCS inventory makeup from SIT and the FLEX pump, the decreased core level is also restored, as shown in Figure A-17. The core is not uncovered and fuel integrity is preserved. The cladding KEPCO & KHNP A3

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 temperature of the active core shown in Figure A-18 also shows that the fuel integrity is maintained during the event.

Figure A-19 shows MSSV and MSADV flow during the event. The MSSV relieves the steam generated by the decay heat of core and thus maintains the SG pressure at a constant value during the hot standby conditions, whereas the MSADV is used for cooling the RCS and maintaining the SG pressure at 1.03 2

kg/cm A (14.7 psia).

3 Figure A-20 shows that the total mass of AFW flow supplied during the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is about 2,839 m (750,000 gal), which is still below the design value of the two AFWSTs inventory.

Therefore, it is concluded that the APR1400 operational strategy for core cooling under simultaneous ELAP and LUHS provides reasonable assurance the appropriate core cooling and fuel integrity during the BDBEE.

The plant can be maintained at hot standby condition without operator action during Phase 1 (0 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the BDBEE). During the phase 2 (8 to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ), the plant can be cooled down to hot shutdown or cold shutdown state depending on the type of coping strategy.

The safe shutdown state is maintained with onsite water for feeding SGs and fuel for GTGs until the end of Phase 2, and it is also continued by the water source and fuel supplied from offsite resources in Phase 3.

A.5.1.3 Subcriticality - Full-Power Operation FLEX Strategy During the RCS cooldown by NCC, subcriticality of the core is maintained by not only the reactor scram but also the injection of borated water by the ACP, the primary FLEX pump, and SIT flow. The subcriticality analysis is performed for the contingency plan among the above two types of full-power operation FLEX strategy. The reason is that this case is more conservative than the other case, because the RCS is further cooled down in the contingency plan. Figure A-21 shows that the borated water injection flow provides negative reactivity sufficient to maintain the core at subcriticality in spite of the positive reactivity feedback due to the moderator temperature coefficient (MTC) and Doppler coefficient.

Figure A-22 shows the reactivity behavior in the case of no RCP leakage, which is a worse condition with respect to criticality. Even in this case, the borated water injection flow also provides negative reactivity sufficient to maintain the core at subcriticality.

A.5.2 Lower-Power Operation and Shutdown Conditions with SGs Available The APR1400 FLEX strategies and their coping capabilities against ELAP concurrent with LUHS during low-power operations and shutdown conditions are described for the following two categories of operation modes.

Operation mode 1 through 3 with SGs available

Operation mode 4 and mode 5 with SGs available A.5.2.1 Coping Capability - Mode 1 through Mode 3 The NCC analysis result for ELAP and LUHS at full power is still valid for the mode 1 through 3 operations, i.e., lower-power operation, startup, and hot standby conditions, because it covers various states of the plant including full-power operation through hot shutdown condition. Therefore, the same FLEX strategy as in the full-power operation can be also applied to the mode 1 through mode 3 KEPCO & KHNP A4

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 operations.

A.5.2.2 Coping Capability - Mode 4 and Mode 5 with SGs Available In the operation mode 4 and 5 with SGs available, the SCS normally maintains the RCS between 176.67 °C(350 °F) (hot shutdown) and 54.44 °C (130 °F ) (cold shutdown), while the SGs are still available. If the event of ELAP concurrent with LUHS occurs during these operation modes, the RCS is heated up and pressurized due to loss of residual heat removal function of the SCS.

If the RCS temperature is initially below the maximum RCS temperature requiring the LTOP, i.e.,

136.11 °C (277 °F), the RCS pressure can be maintained below the LTOP limiting pressure of 43.94 2 2 kg/cm A (625 psia) (20 percent of the RCS hydraulic test pressure of 219.71 kg/cm A [3,125 psia]),

because the LTOP relief valve installed in the SCS automatically opens at the opening setpoint (38.30 2

kg/cm A [530 psig]).

After the RCS temperature increases to the LTOP disable temperature (136.11 °C [277 °F]), the operator tries to isolate the RCS from the SCS, by manually closing the SCS isolation valves. The operator action for isolation of the SCS is finished before the RCS temperature exceeds the SCS entry temperature (176.67 °C [350 °F]). After that, the RCS overpressurization can be protected by POSRVs.

After closing the SCS isolation valves, the RCS temperature and pressure continue to increase, and eventually the RCS returns to the hot standby condition. Then, the SG side feed-and-bleed operation can start cooling down the RCS, as described for the baseline cooling capability for ELAP and LUHS at full-power operation. Consequently, the full-power FLEX strategy can be also applied after the plant returns to hot standby condition.

Supporting analysis to confirm the above scenario has been performed under the assumption that the RCS has been cooled down to the refueling temperature of 54.44 °C (130 °F ), while RCS pressure is 2

kept at the SCS entry pressure of 31.64 kg/cm A (450 psia). The initial RCS pressure and temperature are selected as a conservative combination in the cold shutdown operation range with respect to LTOP.

The initial pressurizer level is 30 percent, which is the normal operating level during the low-mode operation.

Figure A-23 and A-24 show the analysis result for the RCS pressure and temperature during the event, respectively. After the shutdown cooling pump (SCP) stops at time zero, RCS pressure and temperature increase due to loss of residual heal removal function of SCS. However, the increasing rate of the RCS pressure becomes much slower at around 30 minutes, when the LTOP valve opens to mitigate the low-temperature overpressurization (Figure A-25). On the other hand, RCS temperature continues to increase, and reaches the LTOP disable temperature of 136.11 °C(277 °F) at around 2.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months . It can be seen that 2

RCS pressure is maintained well below the LTOP limiting pressure of 43.94 kg/cm A (625 psia) until the RCS temperature reaches the LTOP disable temperature. RCS pressure increases rapidly again at around 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months when the operator is assumed to isolate the SCS from the RCS.

Based on the analysis result, it is concluded that the RCS overpressurization is well protected by the LTOP valve installed in the SCS, until the SCS is isolated by the operator after the RCS temperature exceeds the LTOP disable temperature. The RCS returns to the hot standby condition (above 176.67 °C

[350 °F]) at around 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following the event, so that the operator can isolate the SCS from the RCS and conduct the cooldown operation according to the full-power FLEX strategy. Although the operator action for RCS cooldown is delayed, the RCS overpressurization is successfully limited by the cyclic opening of POSRVs.

KEPCO & KHNP A5

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table A-1 Sequence of Events for Coping Operation Against ELAP/LUHS Occurred at Full-Power Operation Time Event Case 1 Case 2 (Basic strategy) (Contingency plan)

ELAP occurs 0 minute Reactor trip by low RCP speed trip 0 minute Commence manual RCS cooldown with MSADVs at a rate of 27.78 °C/hour (50 °F/hour) 8.0 hours0 days 0 hours 0 weeks 0 months SITs start to inject borated water into RCS 10.40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months 10.15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months RCS hot-leg temperature reaches 176.67 °C (350 °F) (SCS entry condition) 12.78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months 12.76 hours8.796296e-4 days 0.0211 hours 1.256614e-4 weeks 2.8918e-5 months KEPCO & KHNP A6

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 180 2560 RCS SG-1 160 SG-2 2276 140 1991 Pressure (kg/cm2)

Pressure (psia) 120 1707 100 1422 80 1138 60 853 40 569 20 284 0 0 0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72 Time (hr)

Figure A-1 RCS and Steam Generator Pressure (Basic Strategy) 360 680 340 Average 644 Hot leg(averaged) 320 Cold leg(averaged) 608 300 572 Temperature (oC) Temperature (oF) 280 536 260 500 240 464 220 428 200 392 180 356 160 320 140 284 120 248 100 212 0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72 Time (hr)

Figure A-2 RCS Temperature (Basic Strategy)

KEPCO & KHNP A7

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 300 661 SIT-1 SIT-2 250 SIT-3 551 SIT-4 200 441 Flow (kg/sec) Flow (lbm/sec) 150 331 100 220 50 110 0 0

-50 -110 0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72 Time (hr)

Figure A-3 SIT Flow (Basic Strategy) 4.0 8.8 Charging 3.5 Letdown-1 7.7 Letdown-2 Letdown-3 3.0 ACP 6.6 2.5 5.5 Flow (kg/sec) Flow (lbm/sec) 2.0 4.4 1.5 3.3 1.0 2.2 0.5 1.1 0.0 0.0

-0.5 -1.1

-1.0 -2.2 0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72 Time (hr)

Figure A-4 ACP Flow (Basic Strategy)

KEPCO & KHNP A8

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 4.0 8.8 Loop-1a 3.5 Loop-1b 7.7 Loop-2a Loop-2b 3.0 6.6 2.5 5.5 Flow (kg/sec) Flow (lbm/sec) 2.0 4.4 1.5 3.3 1.0 2.2 0.5 1.1 0.0 0.0

-0.5 -1.1

-1.0 -2.2 0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72 Time (hr)

Figure A-5 RCS Leak Flow (Basic Strategy) 110 100 90 80 70 Level (%)

60 50 40 30 20 10 0

-10 0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72 Time (hr)

Figure A-6 RCS Pressurizer Water Level (Basic Strategy)

KEPCO & KHNP A9

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 9 30 Downcomer Core 8 26 7 23 Level (m) 6 20 Level (ft) 5 16 4 13 Top of Active Core 3 10 0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72 Time (hr)

Figure A-7 Collapsed Downcomer and Core Level (Basic Strategy) 400 752 380 716 Hot Rod Bottom Hot Rod 40% height 360 680 Hot Rod 60% height Hot Rod Top 340 644 320 608 Temperature (oC) 300 572 Temperature (oF) 280 536 260 500 240 464 220 428 200 392 180 356 160 320 140 284 120 248 100 212 0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72 Time (hr)

Figure A-8 Cladding Temperature (Basic Strategy)

KEPCO & KHNP A10

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 400 882 ADV-1 Flow ADV-2 Flow 350 MSSV-1 Flow 772 MSSV-2 Flow MSSV-3 Flow 300 MSSV-4 Flow 661 MSSV-5 Flow MSSV-6 Flow 250 551 Flow (kg/sec) Flow (lbm/sec) 200 441 150 331 100 220 50 110 0 0

-50 -110 0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72 Time (hr)

Figure A-9 MSADV, MSSV Flow (Basic Strategy) 3000 792516 2500 660430 Liquid Volume (m3) Liquid Volume (gallons) 2000 528344 1500 396258 1000 264172 500 132086 0 0 0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72 Time (hr)

Figure A-10 Integration of AFW Flow (Basic Strategy)

KEPCO & KHNP A11

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 180 2560 RCS SG-1 160 SG-2 2276 140 1991 120 1707 Pressure (kg/cm2) Pressure (psia) 100 1422 80 1138 60 853 40 569 20 284 0 0 0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72 Time (hr)

Figure A-11 RCS and Steam Generator Pressure (Contingency Plan) 360 680 Average 340 Hot leg(averaged) 644 Cold leg(averaged) 320 608 300 572 280 536 Temperature (oC) Temperature (oF) 260 500 240 464 220 428 200 392 180 356 160 320 140 284 120 248 100 212 0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72 Time (hr)

Figure A-12 RCS Temperature (Contingency Plan)

KEPCO & KHNP A12

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 300 661 SIT-1 SIT-2 250 SIT-3 551 SIT-4 200 441 Flow (kg/sec) Flow (lbm/sec) 150 331 100 220 50 110 0 0

-50 -110 0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72 Time (hr)

Figure A-13 SIT Flow (Contingency Plan) 4.0 8.8 3.5 7.7 3.0 6.6 2.5 5.5 Flow (kg/sec) Flow (lbm/sec) 2.0 4.4 1.5 3.3 1.0 2.2 0.5 1.1 0.0 0.0

-0.5 -1.1

-1.0 -2.2 0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72 Time (hr)

Figure A-14 Primary FLEX Pump Flow (Contingency Plan)

KEPCO & KHNP A13

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 4.0 8.8 Loop-1a 3.5 Loop-1b 7.7 Loop-2a Loop-2b 3.0 6.6 2.5 5.5 Flow (kg/sec) Flow (lbm/sec) 2.0 4.4 1.5 3.3 1.0 2.2 0.5 1.1 0.0 0.0

-0.5 -1.1

-1.0 -2.2 0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72 Time (hr)

Figure A-15 RCS Leak Flow (Contingency Plan) 110 100 90 80 70 Level (%)

60 50 40 30 20 10 0

-10 0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72 Time (hr)

Figure A-16 Pressurizer Water Level (Contingency Plan)

KEPCO & KHNP A14

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 9 30 Downcomer Core 8 26 7 23 Level (m) 6 20 Level (ft) 5 16 4 13 Top of Active Core 3 10 0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72 Time (hr)

Figure A-17 Collapsed Downcomer and Core level (Contingency Plan) 400 752 380 Hot Rod Bottom 716 Hot Rod 40% height 360 680 Hot Rod 60% height Hot Rod Top 340 644 320 608 Temperature (oC) Temperature (oF) 300 572 280 536 260 500 240 464 220 428 200 392 180 356 160 320 140 284 120 248 100 212 0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72 Time (hr)

Figure A-18 Cladding Temperature (Contingency Plan)

KEPCO & KHNP A15

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 400 882 ADV-1 Flow ADV-2 Flow 350 MSSV-1 Flow 772 MSSV-2 Flow MSSV-3 Flow 300 MSSV-4 Flow 661 MSSV-5 Flow MSSV-6 Flow Flow (lbm/sec) 250 551 Flow (kg/sec) 200 441 150 331 100 220 50 110 0 0

-50 -110 0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72 Time (hr)

Figure A-19 MSADV, MSSV Flow (Contingency Plan) 3000 792516 2500 660430 Liquid Volume (m3) Liquid Volume (gallons) 2000 528344 1500 396258 1000 264172 500 132086 0 0 0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72 Time (hr)

Figure A-20 Integration of AFW Flow (Contingency Plan)

KEPCO & KHNP A16

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Reactivity (with RCP Leak) 10 5

0 Reactivity (%)

-5

-10

-15 Total eactivity feedback Modeato Change

-20 Modeato ensity Change Modeato Tempeatue Change Boon

-25 Rod Scam Fuel Tempeatue Change

-30 0 2 4 6 8 10 12 14 16 18 20 22 24 Time (h)

Figure A-21 Reactivity Changes during RCS Cooldown (with RCP Leak) 10 5

0 Reactivity (%)

-5

-10

-15 Total eactivity feedback

-20 Modeato Change Modeato ensity Change Modeato Tempeatue Change Boon

-25 Rod Scam Fuel Tempeatue Change

-30 0 2 4 6 8 10 12 14 16 18 20 22 24 Time (h)

Figure A-22 Reactivity Changes during RCS Cooldown (No RCP Leak)

KEPCO & KHNP A17

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 TS Figure A-23 RCS Pressure TS Figure A-24 RCS Temperature KEPCO & KHNP A18

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 TS Figure A-25 LTOP Valve Flow KEPCO & KHNP A19

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 APPENDIX. B Spent Fuel Pool Time to Boil and Makeup Analysis KEPCO & KHNP Bi

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 TABLE OF CONTENTS B.1 ACCEPTANCE CRITERIA .................................................................................. B1 B.2 KEY ASSUMPTIONS ......................................................................................... B1 B.3 METHODOLOGY ............................................................................................... B2 B.4 EVALUATION RESULTS .................................................................................... B2 LIST OF TABLES Table B-1 Time to Reach SFP Bulk Boiling and Input Parameters ......................................................B4 Table B-2 Time to Reach SFP Water Level 2 and Level 3 ...................................................................B4 Table B-3 Required Makeup Volume and Water Source......................................................................B5 LIST OF FIGURES Figure B-1 Spent Fuel Pool Monitoring Water Level ................................................................................B6 KEPCO & KHNP Bii

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 APPENDIX.B Spent Fuel Pool Time to Boil and Makeup Analysis The APR1400 SFP conditions are analyzed for the BDBEE concurrent with LUHS.

B.1 Acceptance Criteria The following acceptance criteria based on NEI 12-06, Section 3.2.1 are applied to the supporting analysis for the operational strategy for SFP cooling during the ELAP:

Fuel in the SFP remains covered.

B.2 Key Assumptions The SFP time to boil and makeup analysis is performed using the following key assumptions and inputs:

During an ELAP event, spent fuel cooling function by the spent fuel pool cooling system (SFPCS) heat exchangers is lost. Heatup of the SFP and boiling can be credited to cool the spent fuel, provided the water level is maintained above the top of the spent fuel rack.

A conservative number of rack spaces are assumed with all rack spaces.

Heat losses from the SFP are conservatively neglected.

Three conditions analyzed for the SFP decay heat load are:

- SFP decay heat load during mode 1 to mode 4 without full core offload. This is defined as the decay heat generated by one refueling batch offloaded from core after 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months following shutdown, plus the spent fuel assemblies accumulated from the previous refueling operations.

- SFP decay heat load during mode 5 and mode 6 without full core offload.

- SFP decay heat during mode 6 with full core offload (limiting case), which is defined as the decay heat generated by one full core offloaded after 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months following shutdown, plus the spent fuel assemblies accumulated from the previous refueling operations.

Initial SFP temperature is assumed at the maximum as follows:

- SFP decay heat during mode 1 to mode 6 with full core not offloaded: 48.9 (120 °F)

- SFP decay heat during mode 6 with full core offload: 60 (140 °F)

Initial SFP water level is assumed at El 149 ft 0 in (elevation of SFP cleanup suction nozzle).

SFP inventory makeup starts when the water level reaches Level 2 defined in NEI 12-02 and shown in Figure B-1.

Water inventories 3 3

- The water inventory above top of fuel rack: 816.3 m (28,826 ft )

3 3

- Total SFP inventory: 816.3 m (28,826 ft )

KEPCO & KHNP B1

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1

Monitoring water level Based on Figure B-1, the following are the definitions of the water levels:

Level 1: Level adequate to support operation of the normal SFP cooling system Level 2: Level adequate to provide substantial radiation shielding for a person standing on the SFP operating floor Level 3: Level where fuel remains covered, but actions to implement makeup water addition should no longer be deferred B.3 Methodology The SFP time to boil and makeup analysis was performed to determine the bulk SFP heatup time and boiloff rate.

The SFP bulk heatup time is conservatively calculated using t =MCpT/Q, where, M (lbm) is the mass of water in the SFP.

Cp (Btu/lbm°F) is the specific heat.

T (°F) is the temperature rise.

t (hr) is the time to complete the temperature rise.

Q (Btu/hr) is the heat added to the SFP from the spent fuel stored in the pool.

The boiloff rate is calculated using: Boiloff Rate = Q / hfg, where, Q (Btu/hr) is the heat added to the SFP from the spent fuel stored in the pool.

hfg (Btu/lbm) is the latent heat of evaporation.

B.4 Evaluation Results The evaluation results for the SFP time to boil and makeup water analysis are summarized in Tables B-1 through B-3. Based on Tables B-1 and B-2, the following inference can be drawn:

For modes 1 to mode 6 with full core not offloaded, SFP boiling occurs approximately 5.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after an ELAP. As a result of boiling, SFP level reaches 3.05 m (10 ft) of water above the irradiated fuel assemblies in approximately 33.1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after an ELAP. Within 33.1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , Phase 2 actions should be initiated to provide makeup water to the SFP from the RWT using the portable pump.

For mode 6 with maximum SFP heat loads due to a full core offload, the time to boil is reduced to 2.0 hours0 days 0 hours 0 weeks 0 months after an ELAP. As a result of boiling, SPF level reaches 3.05 m (10 ft) of water KEPCO & KHNP B2

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 above the irradiated fuel assemblies in approximately 15.36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months after an ELAP. Within 15.36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , Phase 2 actions should be initiated to provide makeup water to the SFP from the RWT using the portable pump.

In a postulated ELAP, the required SFP makeup flow rate to match the boiloff for modes 1 through 6 with core not offloaded is 237.7 L/m (62.79 gpm), and the SFP makeup flow rate for mode 6 with the full core offload is 493.2 L/min (130.3 gpm).

KEPCO & KHNP B3

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table B-1 Time to Reach SFP Bulk Boiling and Input Parameters MODE Mode 1~6 Mode 6 Parameter (with no full core offload) (with full core offload)

Initial water temperature, °C (°F) 48.9 (120) 60 (140) 3 3 Water density, kg/m (lbm/ft ) 958.4 (59.83) 958.4 (59.83) 3 3 Water volume, m (ft ) 816.3 (28,826) 816.3 (28,826)

Specific heat, kcal/kg-°C (Btu/lb-°F) 1.0 (1.0) 1.0 (1.0)

Heat load, MW (MBtu/hr) 8.55 (29.2) 17.75 (60.6)

Temperature increase rate, hr/°C (hr/°F) 0.11 (0.06) 0.05 (0.03)

Temp. difference to boiling, °C (°F) 51.1 (92.00) 40.0 (72.00)

Time to boiling (hours) 5.4 2.0 Table B-2 Time to Reach SFP Water Level 2 and Level 3 MODE Mode 1~6 Mode 6 Parameter (with no full core offload) (with full core offload) 3 3 Density at 100 °C (212 °F), kg/m (lb/ft ) 958.4 (59.83) 958.4 (59.83)

Latent heat of water, kJ/kg (Btu/lb) 2,252 (969) 2,252 (969)

Boiloff rate, kg/hr (lb/hr) 13,669 (30,134) 28,367 (62,539)

Boiloff rate, L/min (gpm) 237.7 (62.79) 493.2 (130.3)

Time to Level 2 (3.05 m [10 ft] above fuel 33.1 15.36 rack top), (hours)

Time to Level 3 (above fuel rack top), (hours) 62.7 29.6 KEPCO & KHNP B4

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table B-3 Required Makeup Volume and Water Source M (Note 1)

Mode 1~6 Mode 6 ODE (with no full core offload ) (with full core offload)

Parameter Water source RWT RWT 3

Total volume, m (gal) 9,993.49 (2,640,000) 9,993.49 (2,640,000)

Available volume for SFP makeup during total coping 3,443 (909,654) 9,993.49 (2,640,000) 3 time, m (gal)

Time for makeup (72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months minus time to 3.05 m 38.95 56.64 Makeup

[10 ft] above fuel during 72 rack top),

hours (hours)

Required makeup volume, 555.46 (146,737) 1,676 (442,883) 3 m (gal)

Time for makeup (11 days minus Makeup time to 3.05 m 230.95 248.64 during 11 [10 ft] above fuel days (264 rack top),

hours) (hours)

Required makeup volume, 3,294 (870,114) 7,359(1,944,138) 3 m (gal)

Total coping 11.4 days 14.7 days Time (Note 1): RWT can be used as the water source for NCC operation through TDAFWP and SFP makeup during modes 1 through 5 operation with SGs available. However, COL applicants are responsible for determining the available volume of SFP makeup and coping time during modes 5 and 6 with SGs unavailable.

KEPCO & KHNP B5

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 156-0 Normal water level (154-0)

Level 1 Low alarm level (153-10)

Level 2 10 ft above fuel rack top (139-8) 10ft Level 3 Fuel rack top (129-8) 114-0 Figure B-1 Spent Fuel Pool Monitoring Water Level KEPCO & KHNP B6

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 APPENDIX. C 480 V and 4.16 kV Mobile GTG Electrical Loadings KEPCO & KHNP Ci

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 APPENDIX.C 480 V and 4.16 kV Mobile GTG Electrical Loadings The tables below provide a detailed estimate of electrical loading for the 480 V and 4.16 kV mobile GTGs as described in Subsection 5.1.2.6.

Table C-1 480 V Mobile GTG Electrical Loadings (1 of 2) TS KEPCO & KHNP C1

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 TS Table C-1 480 V Mobile GTG Electrical Loadings (2 of 2)

KEPCO & KHNP C2

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 TS Table C-2 4.16 kV Mobile GTG Electrical Loads (1 of 6)

KEPCO & KHNP C3

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 TS Table C-2 4.16 kV Mobile GTG Electrical Loads (2 of 6)

KEPCO & KHNP C4

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 TS Table C-2 4.16 kV Mobile GTG Electrical Loads (3 of 6)

KEPCO & KHNP C5

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 TS Table C-2 4.16 kV Mobile GTG Electrical Loads (4 of 6)

KEPCO & KHNP C6

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table C-2 4.16 kV Mobile GTG Electrical Loads (5 of 6) TS KEPCO & KHNP C7

Non-Proprietary Evaluations and Design Enhancements to Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident APR1400-E-P-NR-14005-NP, Rev. 1 Table C-2 4.16 kV Mobile GTG Electrical Loads (6 of 6) TS KEPCO & KHNP C8

ML17114A520

Contents

Text

1.0 INTRODUCTION

7.0 CONCLUSION

8.0 REFERENCES

1.0 INTRODUCTION

1.1 Arrangement

1.2 Qualification

1.4 Accuracy

1.5 Display

7.0 CONCLUSION

8.0 REFERENCES

Navigation menu

ML17114A520

Text

1.0 INTRODUCTION

7.0 CONCLUSION

8.0 REFERENCES

1.0 INTRODUCTION

1.1 Arrangement

1.2 Qualification

1.4 Accuracy

1.5 Display

7.0 CONCLUSION

8.0 REFERENCES

Navigation menu

Search