ML18079A429

Korea Hydro-Nuclear Power Co., Ltd (Design Control Document) Rev.2 - Tier2 Chapter01-Introduction and General Description of the Plant

ML18079A429
Person / Time
Site:	05200046
Issue date:	03/08/2018
From:	Kim H Korea Hydro & Nuclear Power Co, Ltd
To:	Office of New Reactors
Ward W
References
KOREAHYDRONUC, KOREAHYDRONUC.SUBMISSION.8, APR1400.DCD.NS, APR1400.DCD.NS.3
Download: ML18079A429 (340)

Text

APR1400 DESIGN CONTROL DOCUMENT TIER 2 CHAPTER 1 INTRODUCTION AND GENERAL DESCRIPTION OF THE PLANT APR1400-K-X-FS-14002-NP REVISION 2 FEBRUARY 2018

2018 KOREA ELECTRIC POWER CORPORATION KOREA HYDRO & NUCLEAR POWER CO., LTD All Rights Reserved 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.

Rev. 2

APR1400 DCD TIER 2 CHAPTER 1 - INTRODUCTION AND GENERAL DESCRIPTION OF THE PLANT TABLE OF CONTENTS NUMBER TITLE PAGE CHAPTER 1 - INTRODUCTION AND GENERAL DESCRIPTION OF THE PLANT ................................................................................................... 1.1-1 1.1 Introduction.......................................................................................................... 1.1-1 1.1.1 Plant Location.......................................................................................... 1.1-1 1.1.2 Containment Type ................................................................................... 1.1-1 1.1.3 Reactor Type ........................................................................................... 1.1-1 1.1.4 Power Output ........................................................................................... 1.1-2 1.1.5 Schedule .................................................................................................. 1.1-2 1.1.6 Format and Content ................................................................................. 1.1-2 1.1.6.1 NRC Regulatory Guide 1.206 ............................................... 1.1-2 1.1.6.2 Standard Review Plan ........................................................... 1.1-2 1.1.6.3 Text, Tables, and Figures ...................................................... 1.1-2 1.1.6.4 Page Numbering .................................................................... 1.1-2 1.1.6.5 Proprietary Information ......................................................... 1.1-3 1.1.6.6 Acronyms and Abbreviations ................................................ 1.1-3 1.1.6.7 Amendments .......................................................................... 1.1-3 1.1.7 Combined License Information ............................................................... 1.1-3 1.1.8 References ............................................................................................... 1.1-4 1.2 General Plant Description ................................................................................... 1.2-1 1.2.1 Principal Design Criteria, Operating Characteristics, and Safety Considerations ......................................................................................... 1.2-1 1.2.1.1 Principal Design Objectives .................................................. 1.2-1 1.2.1.2 Fundamental Design Approach ............................................. 1.2-3 1.2.2 Principal Site Characteristics................................................................... 1.2-4 1.2.3 Nuclear Steam Supply System Summary................................................ 1.2-5 1.2.3.1 Reactor................................................................................... 1.2-5 1.2.3.2 Reactor Coolant System and Connecting System ................. 1.2-6 i Rev. 2

APR1400 DCD TIER 2 1.2.4 Engineered Safety Features ..................................................................... 1.2-7 1.2.4.1 Reactor Containment Building .............................................. 1.2-7 1.2.4.2 Safety Injection System ......................................................... 1.2-8 1.2.4.3 Auxiliary Feedwater System ................................................. 1.2-9 1.2.4.4 Containment Spray System ................................................. 1.2-10 1.2.4.5 Containment Isolation System............................................. 1.2-11 1.2.4.6 Engineered Safety Features Filter Systems ......................... 1.2-11 1.2.5 Instrumentation and Control .................................................................. 1.2-12 1.2.5.1 Reactor Trip System ............................................................ 1.2-12 1.2.5.2 Engineered Safety Features System .................................... 1.2-13 1.2.5.3 Component Control System ................................................ 1.2-13 1.2.5.4 Diverse Protection System .................................................. 1.2-13 1.2.5.5 Reactor Control Systems ..................................................... 1.2-14 1.2.5.6 Nuclear Instrumentation ...................................................... 1.2-15 1.2.5.7 Process Monitoring Systems ............................................... 1.2-16 1.2.6 Human-System Interface System .......................................................... 1.2-16 1.2.6.1 Main Control Room............................................................. 1.2-16 1.2.6.2 Remote Shutdown Room..................................................... 1.2-17 1.2.6.3 Qualified Indication and Alarm System .............................. 1.2-17 1.2.6.4 Information Processing System ........................................... 1.2-17 1.2.7 Electrical System ................................................................................... 1.2-18 1.2.8 Steam and Power Conversion System ................................................... 1.2-19 1.2.8.1 Turbine Generator ............................................................... 1.2-19 1.2.8.2 Main Steam System ............................................................. 1.2-20 1.2.8.3 Condensate and Feedwater System ..................................... 1.2-21 1.2.9 Heating, Ventilation, and Air Conditioning System ............................. 1.2-21 1.2.10 Fuel Handling and Storage .................................................................... 1.2-22 1.2.10.1 Fuel Handling ...................................................................... 1.2-22 1.2.10.2 Fuel Storage ......................................................................... 1.2-23 1.2.11 Cooling Water Systems ......................................................................... 1.2-23 1.2.11.1 Circulating Water System ................................................... 1.2-23 1.2.11.2 Essential Service Water System .......................................... 1.2-23 1.2.11.3 Component Cooling Water System ..................................... 1.2-24 1.2.11.4 Chilled Water System .......................................................... 1.2-25 ii Rev. 2

APR1400 DCD TIER 2 1.2.11.5 Spent Fuel Pool Cooling and Cleanup System .................... 1.2-25 1.2.12 Auxiliary Systems ................................................................................. 1.2-26 1.2.12.1 Shutdown Cooling System .................................................. 1.2-26 1.2.12.2 Chemical and Volume Control System ............................... 1.2-26 1.2.12.3 Primary Sampling System ................................................... 1.2-27 1.2.12.4 Condensate Polishing System ............................................. 1.2-28 1.2.12.5 Steam Generator Blowdown System ................................... 1.2-28 1.2.12.6 Compressed Air and Gas Systems....................................... 1.2-29 1.2.12.7 Equipment and Floor Drainage System............................... 1.2-29 1.2.12.8 Fire Protection Program ...................................................... 1.2-30 1.2.12.9 Communication Systems ..................................................... 1.2-30 1.2.12.10 Lighting System .................................................................. 1.2-30 1.2.12.11 Emergency Diesel Generator System .................................. 1.2-31 1.2.12.12 Gas Turbine Generator Facility ........................................... 1.2-31 1.2.12.13 Domestic Water and Sanitary System ................................. 1.2-31 1.2.13 Radioactive Waste Management Systems ............................................. 1.2-32 1.2.13.1 Liquid Waste Management System ..................................... 1.2-32 1.2.13.2 Gaseous Waste Management System .................................. 1.2-33 1.2.13.3 Solid Waste Management System ....................................... 1.2-34 1.2.14 Plant Arrangement Summary ................................................................ 1.2-35 1.2.14.1 Reactor Containment Building ............................................ 1.2-36 1.2.14.2 Auxiliary Building............................................................... 1.2-36 1.2.14.3 Turbine Generator Building ................................................ 1.2-38 1.2.14.4 Compound Building ............................................................ 1.2-39 1.2.14.5 Emergency Diesel Generator Building................................ 1.2-39 1.2.14.6 Alternate Alternating Current Gas Turbine Generator Building ............................................................................... 1.2-39 1.2.14.7 Essential Service Water Building ........................................ 1.2-40 1.2.14.8 Component Cooling Water Heat Exchanger Building ........ 1.2-40 1.2.14.9 Security Buildings ............................................................... 1.2-40 1.2.14.10 Storage Tanks ...................................................................... 1.2-40 1.2.15 Combined License Information ............................................................. 1.2-41 1.2.16 References ............................................................................................. 1.2-41 1.3 Comparison with Other Facilities ...................................................................... 1.3-1 iii Rev. 2

APR1400 DCD TIER 2 1.3.1 Comparison with Similar Facility Designs ............................................. 1.3-1 1.4 Identification of Agents and Contractors .......................................................... 1.4-1 1.4.1 Applicant - Program Manager ................................................................ 1.4-1 1.4.2 Architect Engineer (A/E) - KEPCO E&C .............................................. 1.4-2 1.4.3 Major Equipment Supplier - DOOSAN ................................................. 1.4-2 1.4.4 Nuclear Steam Supply System Designer - KEPCO E&C ...................... 1.4-3 1.4.5 Nuclear Fuel Design and Manufacturing - KEPCO NF ......................... 1.4-3 1.4.6 Combined License Information ............................................................... 1.4-3 1.5 Requirements for Additional Technical Information ...................................... 1.5-1 1.5.1 Fluidic Device Design ............................................................................. 1.5-1 1.5.2 Pilot Operated Safety Relief Valve Design ............................................. 1.5-2 1.5.3 Direct Vessel Injection ............................................................................ 1.5-3 1.5.4 Instrumentation and Control System ....................................................... 1.5-3 1.5.5 References ............................................................................................... 1.5-4 1.6 Material Referenced ............................................................................................ 1.6-1 1.7 Drawings and Diagrams ...................................................................................... 1.7-1 1.7.1 Electrical, Instrumentation, and Control Drawings ................................. 1.7-1 1.7.2 Flow Diagrams ........................................................................................ 1.7-1 1.8 Interfaces with Standard Designs....................................................................... 1.8-1 1.8.1 Combined License Information ............................................................... 1.8-1 1.8.2 References ............................................................................................... 1.8-2 1.9 Conformance with Regulatory Criteria............................................................. 1.9-1 1.9.1 Conformance with Regulatory Guides .................................................... 1.9-1 1.9.2 Conformance with Standard Review Plan............................................... 1.9-2 1.9.3 Generic Issues.......................................................................................... 1.9-2 1.9.4 Operational Experience (Generic Communications) .............................. 1.9-2 1.9.5 Advanced and Evolutionary Light-Water Reactor Design Issues ........... 1.9-3 1.9.6 Conformance with Post-Fukushima NRC Recommendations and Requirements ........................................................................................... 1.9-3 iv Rev. 2

APR1400 DCD TIER 2 1.9.7 Part 21 Notification of Failure to Comply or Existence of a Defect and Its Evaluation ........................................................................ 1.9-4 1.9.8 Combined License Information ............................................................... 1.9-4 1.9.9 References ............................................................................................... 1.9-4 v Rev. 2

APR1400 DCD TIER 2 LIST OF TABLES NUMBER TITLE PAGE Table 1.3-1 Comparison of NSSS Components ........................................................ 1.3-2 Table 1.3-2 Comparison of Plant Components Other Than NSSS.......................... 1.3-15 Table 1.6-1 List of Topical Reports........................................................................... 1.6-2 Table 1.6-2 List of Technical Reports ....................................................................... 1.6-3 Table 1.7-1 Safety-Related Electrical, Instrumentation, and Control Drawings ................................................................................................ 1.7-2 Table 1.7-2 APR1400 System Flow Diagrams ......................................................... 1.7-6 Table 1.8-1 Index of System, Structure, or Component Interface Requirements for APR1400 ................................................................... 1.8-3 Table 1.8-2 Combined License Information Items .................................................... 1.8-5 Table 1.9-1 APR1400 Conformance with Regulatory Guides .................................. 1.9-6 Table 1.9-2 APR1400 Conformance with the Standard Review Plan ..................... 1.9-41 Table 1.9-3 APR1400 Conformance with Generic Issues (NUREG-0933) ............ 1.9-73 Table 1.9-4 APR1400 Conformance with Additional TMI-Related Requirements (10 CFR 50.34(f)) ......................................................... 1.9-77 Table 1.9-5 Generic Communications Applicability to APR1400 .......................... 1.9-87 Table 1.9-6 Summary of SECY Documents Provided in Section C.I.1.9.5 of NRC RG 1.206 ................................................................................. 1.9-88 Table 1.9-7 Conformance with SECY-93-087 ........................................................ 1.9-90 Table 1.9-8 APR1400 Strategies for Addressing Tier 1, 2 and 3 NTTF Recommendations ................................................................................ 1.9-94 vi Rev. 2

APR1400 DCD TIER 2 LIST OF FIGURES NUMBER TITLE PAGE Figure 1.2-1 Typical APR1400 Site Arrangement Plan ............................................ 1.2-43 Figure 1.2-2 General Arrangement Reactor Containment Building Section A-A .......................................................................................... 1.2-44 Figure 1.2-3 General Arrangement Reactor Containment Building Section B-B .......................................................................................... 1.2-45 Figure 1.2-4 General Arrangement Reactor Containment Building El. 69-0 and El. 78-0 ........................................................................ 1.2-46 Figure 1.2-5 General Arrangement Reactor Containment Building El. 100-0 ....... 1.2-47 Figure 1.2-6 General Arrangement Reactor Containment Building El. 114-0 ....... 1.2-48 Figure 1.2-7 General Arrangement Reactor Containment Building El. 136-6 ....... 1.2-49 Figure 1.2-8 General Arrangement Reactor Containment Building El. 156-0 ....... 1.2-50 Figure 1.2-9 General Arrangement Auxiliary Building Section A-A ....................... 1.2-51 Figure 1.2-10 General Arrangement Auxiliary Building Section B-B ....................... 1.2-52 Figure 1.2-11 General Arrangement Auxiliary Building El. 55-0 ............................ 1.2-53 Figure 1.2-12 General Arrangement Auxiliary Building El. 68-0 and El. 86-0 ............................................................................................... 1.2-54 Figure 1.2-13 General Arrangement Auxiliary Building El. 78-0 ............................ 1.2-55 Figure 1.2-14 General Arrangement Auxiliary Building El. 100-0 .......................... 1.2-56 Figure 1.2-15 General Arrangement Auxiliary Building El. 120-0 .......................... 1.2-57 Figure 1.2-16 General Arrangement Auxiliary Building El. 137-6 .......................... 1.2-58 Figure 1.2-17 General Arrangement Auxiliary Building El. 156-0 .......................... 1.2-59 Figure 1.2-18 General Arrangement Auxiliary Building El. 174-0 .......................... 1.2-60 Figure 1.2-19 General Arrangement Auxiliary Building Roof El. 195-0 ................. 1.2-61 Figure 1.2-20 General Arrangement EDG Building El. 63-0, Sections A-A and B-B ................................................................................................. 1.2-62 Figure 1.2-21 General Arrangement EDG Building El. 100-0 and El. 121-6........ 1.2-63 Figure 1.2-22 General Arrangement EDG Building El. 135-0 ................................. 1.2-64 vii Rev. 2

APR1400 DCD TIER 2 Figure 1.2-23 General Arrangement Compound Building Sections A-A and B-B ....................................................................................................... 1.2-65 Figure 1.2-24 General Arrangement Compound Building El. 63-0 ......................... 1.2-66 Figure 1.2-25 General Arrangement Compound Building El. 77-0 ......................... 1.2-67 Figure 1.2-26 General Arrangement Compound Building El. 85-0 ......................... 1.2-68 Figure 1.2-27 General Arrangement Compound Building El. 100-0 ....................... 1.2-69 Figure 1.2-28 General Arrangement Compound Building El. 120-0 ....................... 1.2-70 Figure 1.2-29 General Arrangement Compound Building El. 139-6 ....................... 1.2-71 Figure 1.2-30 General Arrangement Compound Building Roof El. 156-0 .............. 1.2-72 Figure 1.2-31 General Arrangement Turbine Generator Building Section A-A ......... 1.2-73 Figure 1.2-32 General Arrangement Turbine Generator Building Section B-B ......... 1.2-74 Figure 1.2-33 General Arrangement Turbine Generator Building El. 73-0 ............. 1.2-75 Figure 1.2-34 General Arrangement Turbine Generator Building El. 100-0 ........... 1.2-76 Figure 1.2-35 General Arrangement Turbine Generator Building El. 136-6 ........... 1.2-77 Figure 1.2-36 General Arrangement Turbine Generator Building El. 170-0 ........... 1.2-78 Figure 1.2-37 General Arrangement Turbine Generator Building Roof Plan ............. 1.2-79 Figure 1.2-38 AAC Gas Turbine Generator Building Section A-A ............................ 1.2-80 Figure 1.2-39 AAC Gas Turbine Generator Building Plan El. 79-0 and El. 85-6 ............................................................................................... 1.2-81 Figure 1.2-40 AAC Gas Turbine Generator Building Plan El. 100-0 ...................... 1.2-82 Figure 1.2-41 AAC Gas Turbine Generator Building Plan El. 120-0 ...................... 1.2-83 Figure 1.2-42 General Arrangement ESW/CCW Hx Building EL. 81-0 (DIV. I) ................................................................................................. 1.2-84 Figure 1.2-43 General Arrangement ESW/CCW Hx Building EL. 100-0 (DIV. I) ................................................................................................. 1.2-85 Figure 1.2-44 General Arrangement ESW/CCW Hx Building Roof Plan (DIV. I) ................................................................................................. 1.2-86 Figure 1.2-45 General Arrangement ESW/CCW Hx Building Section (DIV. I) ........ 1.2-87 Figure 1.2-46 General Arrangement ESW/CCW Hx Building EL. 81-0 (DIV. II) ................................................................................................ 1.2-88 viii Rev. 2

APR1400 DCD TIER 2 Figure 1.2-47 General Arrangement ESW/CCW Hx Building EL. 10-0 (DIV.II) ................................................................................................. 1.2-89 Figure 1.2-48 General Arrangement ESW/CCW Hx Building Roof Plan (DIV. II) ................................................................................................ 1.2-90 Figure 1.2-49 General Arrangement ESW/CCW Hx Building Section (DIV.II) ................................................................................................. 1.2-91 Figure 1.7-1 Flow Diagram Symbols and Legend ...................................................... 1.7-9 ix Rev. 2

APR1400 DCD TIER 2 ACRONYM AND ABBREVIATION LIST AAC alternate alternating current AAFAS alternate auxiliary feedwater actuation signal AB auxiliary building ABCAEES auxiliary building controlled area emergency exhaust system ABD abnormal blow down AC alternating current ACC analysis computer cabinet ACI american concrete institute ACP auxiliary charging pump ACR advanced control room ACU air cleaning unit ADV atmospheric dump valve AE architect engineer AEA Atomic Energy Act AEB Atomic Energy Bureau AF auxiliary feedwater AFAS auxiliary feedwater actuation signal AFW auxiliary feedwater AFWS auxiliary feedwater system AFWST auxiliary feedwater storage tank AHU air handling unit AI analog input AICC Adiabatic Isochoric Complete Combustion AISC American Institute of Steel Construction AISI American Iron and Steel Institute ALARA as low as is reasonably achievable ALI annual limit on intake ALMS acoustic leak monitoring system ALWR advanced light water reactor x Rev. 2

APR1400 DCD TIER 2 AM accident management AMCA Air Movement and Control Association AMI accident monitoring instrumentation AMS Aerospace Material Specification ANS American Nuclear Society ANSI American National Standards Institute AOC averted off-site property damage costs AOE averted occupational exposures AOO anticipated operational occurrence AOP abnormal operating procedure AOSC averted on-site costs AOV air operated valve APC auxiliary process cabinet APC-S auxiliary process cabinet-safety APD amplified probability distribution APE averted public exposure API American Petroleum Institute APR Advanced Power Reactor APR1400 Advanced Power Reactor 1400 APSD auto-power spectral density APWR advanced pressurized water reactor ARI Air-Conditioning and Refrigeration Institute ARM annunciator response model ARMS area radiation monitoring system ARO 1) all rod out

2) additional reactor operator ART adjusted reference temperature AS 1) accident sequence analysis

2) auxiliary steam ASCE American Society of Civil Engineers ASD alternate shutdown xi Rev. 2

APR1400 DCD TIER 2 ASEP accident sequence evaluation program ASHRAE American Society of Heating, Refrigeration, and Air-Conditioning Engineers ASI axial shape index ASIC Application Specific Integrated Circuit ASME American Society of Mechanical Engineers AST alternative source term ASTM American Society of Testing and Materials ATP authorization to proceed ATS automatic turbine startup ATWS anticipated transient without scram AUC alarm unit cabinet AUX auxiliary AVT all volatile treatment AWP automatic withdrawal prohibit AWS American Welding Society BABT boric acid batching tank BABT boric acid bating tank BAC boric acid concentrator BAMP boric acid makeup pump BAST boric acid storage tank BDAS boron dilution alarm system BDBE beyond design basis event BDBEE beyond design basis external events BDD binary decision diagram BDS blowdown subsystem BHP brake horsepower BISI bypassed and inoperable status indication BLOP bearing lift oil pump BLPB branch line pipe break BM boronometer xii Rev. 2

APR1400 DCD TIER 2 BMT basemat melt through BOC beginning of cycle BOL beginning of life BOP balance of plant BP bistable processor BRL Ballistic Research Laboratory BTP branch technical position BWR boiling water reactor C&L closing and latching CAM continuous air monitor CAP corrective action program CAR corrective action request CAREM code-accuracy-based realistic evaluation method CAS 1) compressed air system

2) central alarm station CAT construction acceptance test CAV cumulative absolute velocity CBD continuous blowdown CBDTM cause-based decision tree methodology CBP computer based procedure CBV cation bed ion exchanger vessel CC component cooling water CCDP conditional core damage probability CCF common - cause failure CCFP conditional containment failure probability CCG control channel gateway CCL component control logic CCP centrifugal charging pump CCS component control system CCTV closed-circuit television CCW component cooling water xiii Rev. 2

APR1400 DCD TIER 2 CCW HX component cooling water heat exchanger CCWLLSTAS component cooling water low-low surge tank actuation signal CCWPH component cooling water pump house CCWS component cooling water system CD 1) complete dependence (HRA)

2) condensate system CDF core damage frequency CDI conceptual design information CEA control element assembly CEAC 1) control element assembly calculator

2) CEA calculator CEAE control element assembly ejection CEDE committed effective dose equivalent CEDM control element drive mechanism CEDMCS control element drive mechanism control system CET core exit thermocouple CEUS central and eastern United States CF cavity flooding CF/SPM critical functions / success path monitoring CFF containment failure frequency CFM critical function monitoring CFR code of federal regulations CFS cavity flooding system CGI commercial grade item CHCS containment hydrogen control system CHF critical heat flux CHMS containment hydrogen monitoring system CHR containment heat removal CHRS containment heat removal system CI containment isolation CIAS containment isolation actuation signal xiv Rev. 2

APR1400 DCD TIER 2 CILRT containment integrated leakage rate test CIM component interface module CIP cleaning in place CIS containment isolation system CIV containment isolation valve CLD control logic diagram CLVPS containment low volume purge system CM 1) condition monitoring

2) containment monitoring CMAA Crane Manufacturers Association of America CMTR Certified Material Test Report CNMT containment COE cost of enhancement COL combined license COLA combined license application COLR core operating limits report COLSS core operating limit supervisory system COMP compound COO chief operations officer CP condensate polishing CPC core protection calculator CPCS core protection calculator system CPCSC core protection calculator system cabinet CPIAS containment purge isolation actuation signal CPM control panel multiplexer CPP CEA position processor CPS condensate polishing system CPU central processing unit CR control room CRC cyclical redundancy check CRDS control rod drive system xv Rev. 2

APR1400 DCD TIER 2 CRE control room envelope CREACS control room emergency makeup air cleaning system CREVAS control room emergency ventilation actuation signal CRF carryout rate fraction CRTF central receiver test facility CRX Canadian research reactor CS 1) containment spray

2) core support

3) communication section CSAS containment spray actuation signal CSB core support barrel CSDRS certified seismic design response spectra CSF critical safety function CSHX containment spray heat exchanger CSP containment spray pump CSS containment spray system CST condensate storage tank CT condensate storage and transfer system CTS concentrate treatment system CUF cumulative usage factor CV 1) control valve

2) chemical and volume control system CVAP comprehensive vibration assessment program CVCS chemical and volume control system CVN Charpy V-notch CW circulating water CWP circulating water pump CWS circulating water system CWT chemical waste tank D3CA diversity and defense-in-depth coping analysis DA data analysis xvi Rev. 2

APR1400 DCD TIER 2 DAC 1) derived air concentration

2) design acceptance criteria DAS diverse actuation system DAU data acquisition unit DAW dry active waste DBA design basis accident DBE design basis event DBFL design basis flooding level DBPB design basis pipe break DC 1) direct current

2) design certification DCD Design Control Document DCF 1) dose conversion factor

2) dynamic containment failure DCH direct containment heating DCN design change notice DCN-I data communication network-information DCS distributed control system DDCC drawing and document control center DDE deep dose equivalent DDT deflagration to detonation transition DE dose equivalent DEDLSB double ended discharge leg slot break DEG/PD double-ended guillotine at the pump discharge leg DEHLSB double ended hot leg slot break DELLOC double-ended break of the letdown line outside containment DESLSB double ended suction leg slot break DET decomposition event tree DF decontamination factor DFL dynamic fluid loads DFOT diesel fuel oil tank xvii Rev. 2

APR1400 DCD TIER 2 DG diesel generator DGA diesel generator area DI 1) digital input

2) design implementation DIF dynamic impact factor DIS diverse indication system DIT discrete integral transport DLF dynamic load factor DLM Diffusion Layer Model DMA diverse manual ESF actuation DMDS diagnostic monitoring and display system DN nominal diameter DNB departure from nucleate boiling DNBR departure from nucleate boiling ratio DO 1) dissolved oxygen

2) digital output DOOSAN Doosan Heavy Industries & Construction Co., Ltd DOT U.S. department of transportation DPS diverse protection system DRC dropped rod contact DRCS digital rod control system DS disconnect switch DVI direct vessel injection DW dead weight DWST demineralized water storage tank EAB exclusion area boundary EAC emergency alternating current EBD emergency blowdown EBS estimated break size ECC emergency core cooling ECCS emergency core cooling system xviii Rev. 2

APR1400 DCD TIER 2 ECF early containment failure ECSA electrical conductor sealing assembly ECSBS emergency containment spray backup system ECT Eddy current testing ECWS essential chilled water system EDE effective dose equivalent EDECAIES emergency diesel engine combustion air intake and exhaust system EDECWS emergency diesel engine cooling water system EDEFOS emergency diesel engine fuel oil system EDELS emergency diesel engine lubrication system EDESS emergency diesel engine starting system EDG emergency diesel generator EDGB emergency diesel generator building EDH electric duct heater EDMG extensive damage mitigation guideline EDT equipment drain tank EF 1) error factor

2) engineered safety features actuation system EFDS equipment and floor drainage system EFPD effective full-power day EFPY effective full-power year EHC electro-hydraulic control EJMA Expansion Joint Manufacturers Association ELAP extended loss of ac power EM seismic monitoring EMC electromagnetic compatibility EMI electromagnetic interference ENFMS ex-core neutron flux monitoring system EO electrical operator EOB end of blowdown EOC end of cycle xix Rev. 2

APR1400 DCD TIER 2 EOF emergency operation facility EOL end of life EOP emergency operating procedure EOPR end of post-reflood EOR end of reflood EPA 1) electrical penetration assembly

2) U.S Environmental Protection Agency EPD external pressure differential EPG emergency procedure guideline EPM engineering procedures manual EPRI Electric Power Research Institute EQ environmental qualification EQAP engineering quality assurance procedure ERDS emergency response data system ERF emergency response facility ERVC external reactor vessel cooling ES equipment survivability ESA extension shaft assembly ESCM ESF-CCS soft control module ESF engineered safety features ESF-CCS engineered safety features - component control system ESR electro-hydraulic actuated spring return ESW essential service water ESWPS essential service water pump structure ESWS essential service water system ET event tree ETAP electrical transient analyzer program ETS emergency trip system EVSE ex-vessel steam explosion EWT equipment waste tank FA flame acceleration xx Rev. 2

APR1400 DCD TIER 2 FAC flow-accelerated corrosion FACT fuel assembly compatibility test FAP fuel alignment plate FATT fracture appearance transition temperature FC fully closed FCAW flux cored arc welding FCI fuel-coolant interaction FCR field change request FDS floor drain system FDT 1) floor drain tank

2) functional definition table FEI fluid-elastic instability FEM finite element model FF flash fraction FHA 1) fuel handling accident

2) fuel handling area

3) fire hazards analysis FHAEES fuel handling area emergency exhaust system FHEVAS fuel handling area emergency ventilation actuation signal FHS fuel handling system FIDAS fixed in-core detector amplification system FIRS foundation input response spectra FIV flow-induced vibration FLB feedwater line break FLC factored load category FLEX diverse and flexible coping strategies FME foreign material exclusion FMEA failure modes and effects analysis FO fully open FOM fiber optic modem FOS factor of safety xxi Rev. 2

APR1400 DCD TIER 2 FP fire protection FPD flat panel display FPDIL full-power-dependent insertion limit FPP fire protection plan FPS fire protection system FRA/FA functional requirements analysis and function allocation FS factor of safety FSAR final safety analysis report FSCEA full-strength CEA FSSA fire safe shutdown analysis FT fault tree FTC fuel temperature coefficient FV Fussell-Vesely FW feedwater FWCS feedwater control system FWCV feedwater control valve FWLB feedwater line break FWPB feedwater pipe break GC group controller GCB generator circuit breaker GCP general control procedure GDC general design criteria (of 10 CFR Part 50, Appendix A)

GI gastrointestinal GIB gas insulated bus GIS 1) event-generated iodine spike

2) gas insulated substation GL Generic Letter GMRS ground motion response spectra GOP general operating procedure GOTHIC generation of thermal-hydraulic information for containment GRID-LOOP grid-centered loss of offsite power xxii Rev. 2

APR1400 DCD TIER 2 GRID-SBO grid-centered station blackout GRS gaseous radwaste system GRV gravity roof ventilator GSE gland steam packing exhauster GSERMS gas stripper effluent radiation monitoring system GSI Generic Safety Issue GTAW gas tungsten arc weld GTG gas turbine generator GTRN general transient GWMS gaseous waste management system GWR guided wave radar HA human action HAZ heat-affected zone HCBD high-capacity blowdown HCLPF high confidence of low probability of failure HCOG hydrogen control owners group HCR/ORE human cognitive reliability / operator reliability experiment HD 1) high dependence (HRA)

2) heater drain

3) HSI design HDSR historical data storage and retrieval HE human error HED human engineering discrepancy HEI Heat Exchange Institute HELB high-energy line break HEP human error probability HEPA high-efficiency particulate air HF human factors HFE human factors engineering HFEPP human factors engineering program plan HFP hot full power xxiii Rev. 2

APR1400 DCD TIER 2 HFT hot functional test HG containment hydrogen control system HHAS high-humidity actuation signal HI hydrogen igniter HIC high-integrity container HID high-intensity discharge HIS Hydraulic Institute Standard HJTC heated junction thermocouple HLI hot leg injection HMS hydrogen mitigation system HP high pressure HPCI high-pressure coolant injection HPME high-pressure melt ejection HPPT high pressurizer pressure trip HPS Health Physics Society HPSC high-pressure seal cooler HRA human reliability analysis HRAS high radiation actuation signal HRHF hard rock high frequency HRR heat release rate HSB hot standby HSD hot shutdown HSGL high steam generator level HSI human-system interface HSIS human-system interface system HSS high safety significance HT high temperature HVAC heating, ventilation, and air conditioning HVT holdup volume tank HX heat exchanger HZP hot zero power xxiv Rev. 2

APR1400 DCD TIER 2 I&C instrumentation and control I/O input/output IA instrument air IAS instrument air system IBA inner barrel assembly ICC inadequate core cooling ICCMS inadequate core cooling monitoring system ICDP incremental core damage probability ICI in-core instrumentation ICR information and control requirement ICRP International Commission on Radiological Protection ID 1) inner diameter

2) identification IE 1) initiating events analysis

2) Inspection and Enforcement IEC International Electrotechnical Commission IED internal effective dose IEEE Institute of Electrical and Electronics Engineers IEPRA internal events probabilistic risk assessment IF internal flooding analysis IFPD information flat panel display IHA 1) integrated head assembly

2) important human action ILRT integrated leak rate test INPO Institute of Nuclear Power Operations INVINJ in-vessel injection IOSGADV inadvertent opening of a steam generator atmospheric dump valve IP implementation plan IPB isolated phase bus IPS information processing system IRSF interim radwaste storage facility xxv Rev. 2

APR1400 DCD TIER 2 IRWST in-containment refueling water storage tank IS internal structure ISA Instrument Society of America ISG Interim Staff Guidance ISI inservice inspection ISLOCA intersystem loss-of-coolant accident ISM independent support motion ISRS in-structure response spectra IST inservice testing ISV 1) integrated system validation

2) intermediate stop valve ITA important to availability ITAAC inspections, tests, analyses, and acceptance criteria ITC isothermal temperature coefficient ITP 1) interface and test processor

2) inspection and test plan ITS 1) issue tracking system

2) important to safety IV intercept valve IVMS internal vibration monitoring system IVSE in-vessel steam explosion IW in-containment refueling water storage system IWPP independent water and power producer IWSS in-containment water storage system IX ion exchange J-R J-resistance JOG Joint Owner Group KAERI Korea Atomic Energy Research Institute KEPCO Korea Electric Power Corporation KEPCO E&C KEPCO Engineering & Construction Co., Inc.

KEPCO NF KEPCO Nuclear Fuel Co., Ltd.

xxvi Rev. 2

APR1400 DCD TIER 2 KHNP Korea Hydro & Nuclear Power Co., Ltd.

KWU Kraftwerk Union AG LAN local area network LASRT low-activity spent resin storage tank LB large break LBB leak before break LBLOCA large-break loss-of-coolant accident LC 1) lock close

2) loop controller

3) load center LCF late containment failure LCL local coincidence logic LCO limiting conditions for operation LCP local control panel LCS local control station LD low dependence (HRA)

LDLB letdown line break LDP large display panel LE LERF Analysis LED light-emitting diode LEL lower electrical limit LERF large early release frequency LFW loss of normal feedwater flow LGS lower group stop LHGR linear heat generation rate LHR linear heat rate LHS Latin hypercube sampling LL large LOCA LLHS light load handling system LMFBR liquid metal cooled fast breeder reactor LO local operator xxvii Rev. 2

APR1400 DCD TIER 2 LOAC loss of nonemergency ac power LOCA loss-of-coolant accident LOCCW loss of component cooling water LOCV loss of condenser vacuum LODC loss of dc power LOESW loss of essential service water LOF 1) left-out-force

2) loss of flow LOFW 1) loss of normal feedwater flow

2) loss of main feedwater LOIA loss of instrument air LOLA loss of large areas LOOP loss of offsite power LP low pressure LPD local power density LPLD low PZR pressure and low DNBR LPMS loose parts monitoring system LPSD low power and shutdown LPZ low population zone LRC locked rotor current LRF large release frequency LRS liquid radwaste system LSGL low steam generator level LSS lower support structure LSSB large secondary side break LSSS limiting safety system setting LST lowest service temperature LT low temperature LTC long-term cooling LTOP low temperature overpressure protection LUHS loss of normal access to ultimate heat sink xxviii Rev. 2

APR1400 DCD TIER 2 LWMS liquid waste management system LWR light water reactor M&E mass and energy MAAP modular accident analysis program MBLOCA medium break loss-of-coolant accident MBV mixed-bed ion exchanger vessel MCA multiple compartment analysis MCC motor control center MCCI molten corium concrete interaction MCL main coolant loop MCR main control room MDNBR minimum departure from the nucleate boiling ratio MDS makeup demineralizer system MELB moderate-energy line break MF membrane filter MFIV main feedwater isolation valve MFLB main feedwater line break MFS main feedwater system MFW main feedwater MG motor- generator MG Set motor-generator set MI minimum inventory ML manufacturing license MMI modified Mercalli intensity MOP main oil pump MORS membrane oxygen removal subsystem MOV motor-operated valve MRP materials reliability program MS main steam MSADV main steam atmospheric dump valve MSADVIV MSADV isolation valve xxix Rev. 2

APR1400 DCD TIER 2 MSE main steam enclosure MSGTR multiple steam generator tube rupture MSIS main steam isolation signal MSIV main steam isolation valve MSIVBV main steam isolation valve bypass valve MSLB main steam line break MSO multiple spurious operation MSPI mitigating systems performance index MSR 1) moisture separator reheater

2) maximum steaming rate MSS main steam system MSSV main steam safety valve MSV main steam valve MSVH main steam valve house MT main transformer MTC moderator temperature coefficient MTP maintenance test panel MUX multiplexer MWD/MTU megawatt-days per metric ton of Uranium NA not applicable NDE nondestructive examination NDRC national defense research council NDTT nil-ductility transition temperature NEC National Electrical Code NEI Nuclear Energy Institute NEM nodal expansion method NEMA National Electrical Manufacturers Association NFPA National Fire Protection Association NFR new fuel rack NI nuclear island NIMS NSSS integrity monitoring system xxx Rev. 2

APR1400 DCD TIER 2 NLO non-licensed operator NNS non-nuclear safety NO normal operation NOP normal operating procedure NP non-Class 1E 13.8 kV auxiliary power system NPCS NSSS process control system NPP nuclear power plant NPS nominal pipe size NPSH net positive suction head NPSHA net positive suction head available NPSHR net positive suction head required NPSS normal primary sampling system NR narrow range NRC United States Nuclear Regulatory Commission NRV 1) non-return check valve

2) net present value NS non-seismic NSA neutron source assembly NSAC Nuclear Safety Analysis Center NSSS nuclear steam supply system NT normal torque NTS Nevada Test Site NTTF near term task force NUREG NRC technical report designation OA operational assessment OBE operating basis earthquake OD outside diameter ODCM offsite dose calculation manual OECD Organization for Economic Cooperation and Development OER operating experience review OFAF oil forced air forced xxxi Rev. 2

APR1400 DCD TIER 2 OHLHS overhead heavy load handling system OM operator module ONAF oil natural air forced ONAN oil natural air natural OPR Optimized Power Reactor ORNL Oak Ridge National Laboratory OSC operational support center P-CCS process-component control system P-T Limit pressure-temperature limitation P&ID piping and instrumentation diagram PA public address PABX private automatic branch telephone exchange PACU packaged air conditioning unit PAL personnel air lock PAR passive autocatalytic recombiner PASS post-accident sampling system PAT power ascension test PAU physical analysis unit PBX plant telephone exchange PC prime contractor PCA primary coolant activity PCB power circuit breaker PCMI pellet cladding mechanical interaction PCP project control procedure PCS power control system PCT peak cladding temperature PCWS plant chilled water system PDIL power-dependent insertion limit PDS plant damage state PED piping evaluation diagram PERMSS process and effluent radiation monitoring and sampling system xxxii Rev. 2

APR1400 DCD TIER 2 PF 1) penalty factor

2) 4.16 kV Class 1E auxiliary power PGA peak ground acceleration PI pressure indicator PIS pre-accident iodine spike PIV pressure isolation valve PLC programmable logic controller PLCS pressurizer level control system PLHGR peak linear heat generation rate PLM priority logic module PLOCCW partial loss of component cooling water PLOESW partial loss of essential service water PMF probable maximum flood PMP probable maximum precipitation PMWP probable maximum winter precipitation PNS permanent non-safety POL power operating limit PORV power-operated relief valve POS plant operational state POSRV pilot operated safety relief valve POV power-operated valve PPCS pressurizer pressure control system PPM project procedures manual PPS 1) plant protection system

2) preferred power supply PPSC plant protection system cabinet PRA probabilistic risk assessment PRCSCD RCS pressure at the time of core damage PRM process radiation monitor PRMS process radiation monitoring system PRV process representative value xxxiii Rev. 2

APR1400 DCD TIER 2 PSA probabilistic safety assessment PSAR preliminary safety analysis report PSCEA part-strength control element assembly PSD power spectral density PSHA probabilistic seismic hazard analysis PSI preservice inspection PSR pneumatically actuated spring return PSW primary shield wall PTC peak cladding temperature PTLR pressure and temperature limits report PTS 1) pressurized thermal shock

2) primary to secondary PV preliminary validation PVNGS Palo Verde nuclear generating station PVRC Pressure Vessel Research Committee PWR pressurized water reactor PWSCC primary water stress corrosion cracking PX primary sampling system PZR pressurizer QA quality assurance QAP quality assurance procedure QAPD quality assurance program description QIAS qualified indication and alarm system QIAS-N qualified indication and alarm system - non-safety QIAS-P qualified indication and alarm system - p QU quantification R/O reverse osmosis RADTRAD Radionuclide Transport, Removal, and Dose RAM random access memory RAP reliability assurance program RAW risk achievement worth xxxiv Rev. 2

APR1400 DCD TIER 2 RB reactor building RC reactor coolant RCA radiologically contolled area RCB reactor containment building RCC remote control center RCCA rod cluster control assembly RCFC reactor containment fan cooler RCGV reactor coolant gas vent RCGVS reactor coolant gas vent system RCIC reactor core isolation cooling RCL reactor coolant loop RCP reactor coolant pump RCPB reactor coolant pressure boundary RCPS reactor power cutback system RCPSSSS reactor coolant pump shaft speed sensing system RCPVMS reactor coolant pump vibration monitoring system RCS reactor coolant system RCY reactor critical-year RD rapid depressurization RDS radioactive drain system RDT reactor drain tank REMP Radiological and Environmental Monitoring Program RFI request for information RFI radio frequency interference RG Regulatory Guide RHR residual heat removal RIA reactivity-initiated accident RIHA risk-important human action RLE review level earthquake RLS radioactive laundry subsystem RMI reflective metal insulation xxxv Rev. 2

APR1400 DCD TIER 2 RMS radiation monitoring system RMTS risk-managed technical specifications RMWT reactor makeup water tank RO reactor operator RP reactor protection RPCB reactor power cutback RPCS reactor power cutback system RPS reactor protection system RPV reactor pressure vessel RRS required response spectra RSC remote shutdown console RSF RCP seal LOCA RSG rapid ex-vessel steam generation RSPT reed switch position transmitter RSR remote shutdown room RSSH resin sluice supply header RT reactor trip RTCB reactor trip circuit breaker RTD resistance temperature detector RTE random turbulent excitation RTNDT reference temperature for nil-ductility transition RTNSS regulatory treatment of non-safety systems RTO reactor trip override RTOTT reactor trip on turbine trip RTP 1) rated thermal power

2) return to power RTPTS reference temperature (pressurized thermal shock)

RTS reactor trip system RTSG reactor trip switchgear RTSS reactor trip switchgear system RV reactor vessel xxxvi Rev. 2

APR1400 DCD TIER 2 RVI reactor vessel internals RVLMS reactor vessel level monitoring system RVR reactor vessel rupture RVUH reactor vessel upper head RWP radiation work permit RWT raw water tank RY reactor-year S&Q staffing and qualification SAFDL specified acceptable fuel design limit SAM seismic anchor movement SAMA severe accident mitigation alternative SAMDA severe accident mitigation design alternative SAMG severe accident management guideline SAS service air system SAT standby auxiliary transformer SAW submerged arc welding SBCS steam bypass control system SBLOCA small-break loss-of-coolant accident SBO station blackout SC 1) shutdown cooling

2) safety console

3) safety critical SCC stress corrosion cracking SCETCh severe combined environment test chamber SCP shutdown cooling pump SCS shutdown cooling system SCU statistical combination of uncertainties SDC shutdown cooling SDCHX shutdown cooling heat exchanger SDL serial data link SDM shutdown margin xxxvii Rev. 2

APR1400 DCD TIER 2 SDN safety system data network SECY Secretary of the Commission, Office of the NRC SER safety evaluation report SF 1) stratified flow

2) single failure SFD spent fuel damage SFG structural fill granular SFHM spent fuel handling machine SFP spent fuel pool SFPCCS spent fuel pool cooling and cleanup system SFPCL SFP cleanup loop SFR spent fuel rack SG steam generator SGBS steam generator blowdown system SGI safeguard information SGMSR steam generator maximum steaming rate SGTR steam generator tube rupture SI safety injection SI units International System of Units SIAS safety injection actuation signal SIF stress intensification factor SIFT safety injection filling tank SIP safety injection pump SIRCP startup of an inactive reactor coolant pump SIS safety injection system SIT 1) safety injection tank

2) structural integrity test SIT-FD safety injection tank with fluidic device SKN Shin-Kori nuclear power plant SL surge line SLB steam line break xxxviii Rev. 2

APR1400 DCD TIER 2 SLBFP large steam line break during full-power operation SLBZP large steam line break during zero-power operation SMA seismic margin analysis SMACNA Sheet Metal and Air Conditioning Contractors National Association SMAW shielded metal arc weld SMS seismic monitoring system SODP shutdown overview display panel SOE sequence of events SOP system operating procedure SOV solenoid-operated valve SPADES+ safety parameter display and evaluation system+

SPAR-H standardized plant analysis risk - human reliability SPDS safety parameter display system SPERT special power excursion reactor test program SPM success path monitoring SPND self-powered neutron detector SPTA standard post-trip action SQSDS seismic qualification summary data sheet SRI Stanford Research Institute SRLST Spent Resin Long-tern Storage Tank SRM 1) Staff Requirements Memorandum

2) standard reference material SRO senior reactor operator SRP Standard Review Plan SRS solid radwaste system SRSS square root of the sum of the squares SRST spent resin storage tank SRV safety relief valve SS stainless steel SSA safety shutdown analysis SSC structure, system, or component xxxix Rev. 2

APR1400 DCD TIER 2 SSE safe shutdown earthquake SSI soil-structure interaction SSIE supporting system initiating event SSM saturation margin monitor SSS secondary sampling system SSW secondary shield wall ST stud tensioner STA shift technical advisor STC source term category STP standard temperature & pressure SV suitability verification SWGR switchgear SWMS solid waste management system SWYD switchyard SX essential service water system SY systems analysis T&M test and maintenance T/C reactor inlet temperature, T(cold)

T/G turbine-generator T/H reactor outlet temperature, T(hot)

TA task analysis TAA transient and accident analysis TAVG average temperature TB turbine building TBS turbine bypass system TBV turbine bypass valve TCB trip circuit breaker TCE two-cell equilibrium TCOLD cold leg temperature TCS turbine control system TDAFWP turbine-driven auxiliary feedwater pump xl Rev. 2

APR1400 DCD TIER 2 TDH total dynamic head TDR time domain reflectometry TEDE total effective dose equivalent TEMA Tubular Exchanger Manufacturers Association TEPCO Tokyo Electric Power Company TG turbine generator TGB turbine generator building TGBCCW turbine generator building closed cooling water TGBOCWS turbine generator building open cooling water system TGCS turbine generator control system TSSS turbine steam seal system THD total harmonic distortion THERP technique for human error rate prediction TI temperature indicator TI-SGTR temperature-induced steam generator tube rupture TID 1) technical information document

2) total integrated dose TIHA treatment of important human action TIV temperature isolation valve TLOCCW total loss of component cooling water TLOESW total loss of essential service water TLOFW total loss of feedwater TMI Three Mile Island TNDT nil-ductility transition temperature TO turbine operator TRAN transient TREF reference temperature TRS test response spectrum TS technical specification TSC technical support center TSO transmission system operator xli Rev. 2

APR1400 DCD TIER 2 TSP 1) tri-sodium phosphate

2) transmission system provider TSSS turbine steam seal system TT thermally treated UAT unit auxiliary transformer UEL upper electrical limit UGS upper guide structure UHS ultimate heat sink UHSRS ultimate heat sink related structure UL Underwriters Laboratories UPC ultimate pressure capacity UPS uninterruptible power supply URS uniform response spectrum USE upper-shelf energy USEPA U.S. Environmental Protection Agency USH uniform support motion USI unresolved safety issue UV undervoltage V&V verification and validation Vac voltage alternating current VB vessel breach VBPSS vital bus power supply system VCT volume control tank VD emergency diesel generator area HVAC Vdc voltage direct current VDU visual display unit VEWFDS very early warning fire detection system VFTP ventilation filter testing program VG ESW intake structure/CCW heat exchanger building HVAC VIPER Vibration Investigation and Pressure Drop Experimental Research VK auxiliary building controlled area HVAC xlii Rev. 2

APR1400 DCD TIER 2 VO auxiliary building clean area HVAC VOPT variable overpower trip(signal)

VPN virtual private network VSP variable setpoint VU miscellaneous building HVAC VWO valve wide open WCT waste collection tank WDT watchdog timer WLS wet layup subsystem WO chilled water system WPS welding procedure specification WR wide range WRC Welding Research Council WT turbine generator building closed cooling water system WWTF waste water treatment facility WWTS wastewater treatment system ZD zero dependence (HRA)

ZOI zone of influence ZPA zero-period acceleration xliii Rev. 2

APR1400 DCD TIER 2 CHAPTER 1 - INTRODUCTION AND GENERAL DESCRIPTION OF THE PLANT 1.1 Introduction Korea Hydro & Nuclear Power Co., Ltd. (KHNP) has designed the APR1400, an evolutionary light water reactor (LWR). Korea Electric Power Corporation (KEPCO) and KHNP submit the Design Control Document (DCD) of the APR1400 design for U.S.

Nuclear Regulatory Commission (NRC) review and approval under the provisions of 10 CFR Part 52. KEPCO and KHNP request the issuance of a standard design certification for the APR1400 in accordance with 10 CFR Part 52, Subpart B. This DCD and application for design certification are based on KHNP design experience and the ABB-CE System 80+ certified design.

1.1.1 Plant Location The APR1400 is designed for use at a site with the parameters that are described in Chapter 2 of this DCD. The combined license (COL) applicant that references the APR1400 design certification is to identify the actual plant site location (COL 1.1(1)).

1.1.2 Containment Type The APR1400 containment is a steel-lined prestressed concrete structure that consists of a right circular cylinder with a hemispherical dome on a reinforced concrete common basemat. There is no structural connection between the free-standing portion of the containment and adjacent structures other than penetrations and associated supports. The containment retains integrity at the pressure and temperature conditions associated with the most limiting design basis event (DBE) without exceeding the design leakage rate.

Access to the containment is provided by personnel air locks and an equipment hatch.

Penetrations are also provided for electrical and mechanical components and for the transport of nuclear fuel.

1.1.3 Reactor Type The APR1400 nuclear steam supply system (NSSS) is a KHNP-designed evolutionary two-loop pressurized water reactor (PWR).

1.1-1 Rev. 2

APR1400 DCD TIER 2 1.1.4 Power Output The APR1400 net electrical power output is approximately 1,400 MWe, depending on site conditions. The NSSS rated thermal power is 4,000 MWt with a core thermal power of 3,983 MWt.

1.1.5 Schedule The COL applicant that references the APR1400 is to provide estimated schedules for the completion of construction and the start of commercial operation (COL 1.1(2)).

1.1.6 Format and Content 1.1.6.1 NRC Regulatory Guide 1.206 The format and content of this DCD are based on the guidance in NRC Regulatory Guide (RG) 1.206 (Reference 1). To the extent practical, the chapter, section, subsection, and paragraph headings in the APR1400 DCD are consistent with NRC RG 1.206.

1.1.6.2 Standard Review Plan Preparation of this DCD has followed the guidance in the NRC Standard Review Plan (SRP), NUREG-0800 (Reference 2). An evaluation of the conformance with the SRP is provided in Section 1.9.

1.1.6.3 Text, Tables, and Figures Tables and figures are identified by section or subsection number followed by a sequential number (for example, Table 1.3-1 is the first table of Section 1.3). Tables and figures are placed at the end of the applicable sections immediately following the text. Figures include drawings, graphs, and photographs.

1.1.6.4 Page Numbering Pages are numbered sequentially in each section and are identified by the section number followed by a sequential number, and at the beginning of each chapter.

1.1-2 Rev. 2

APR1400 DCD TIER 2 1.1.6.5 Proprietary Information This document includes no information that is proprietary to KHNP. The portions of this document that are classified as sensitive and will be withheld from the public according to 10 CFR 2.390 are indicated and provided separately to the NRC.

As noted in Section 1.6, the DCD Tier 2 references topical and technical reports that contain proprietary information. In these cases, in Tables 1.6-1 and 1.6-2, the non-proprietary version of the topical or technical report is also identified.

1.1.6.6 Acronyms and Abbreviations The acronyms and abbreviations used in this DCD are provided after the list of figures at the beginning of this chapter.

1.1.6.7 Amendments The APR1400 DCD will be amended, if necessary, as the APR1400 design is finalized.

The DCD will also be amended as a result of the NRC review. To identify the amendments, the following guidelines will be followed:

a. Amended portions will be indicated by vertical lines on the right hand side of the page. The vertical lines will identify only the latest amendment (i.e., amendments since the previous revision number).

b. Figure changes will be indicated by vertical lines on the right hand side of the page. Vertical lines will identify only the latest amendment.

c. Responses to NRC questions will be incorporated, as applicable, into revisions of the DCD.

1.1.7 Combined License Information COL 1.1(1) The COL applicant that references the APR1400 is to identify the actual plant site location.

COL 1.1(2) The COL applicant that references the APR1400 is to provide estimated schedules for the completion of construction and the start of commercial operation.

1.1-3 Rev. 2

APR1400 DCD TIER 2 1.1.8 References

1. Regulatory Guide 1.206, Combined License Applications for Nuclear Power Plants (LWR Edition), U.S. Nuclear Regulatory Commission, June 2007.

2. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (LWR Edition), U.S. Nuclear Regulatory Commission, various dates and revisions.

1.1-4 Rev. 2

APR1400 DCD TIER 2 1.2 General Plant Description This section contains a summary of the principal design criteria, operating characteristics, safety considerations, and major structures and systems. This section also includes a site plan and the general arrangement of major structures and equipment.

The combined license (COL) applicant is to prepare a complete and detailed site plan (COL 1.2(1)).

1.2.1 Principal Design Criteria, Operating Characteristics, and Safety Considerations 1.2.1.1 Principal Design Objectives The following subsection provides the principal design objectives for the safety, reliability, and performance of the plant. These objectives are the basis of the principal design criteria for the APR1400.

1.2.1.1.1 Safety Design Objectives The safety design objectives of the APR1400 are as follows:

a. Simplify plant design and operation, as described in Subsection 1.2.1.2.1.

b. Provide the proper safety margin for a more forgiving and resilient plant, as described in Subsection 1.2.1.2.2.

c. Improve the human-system interface system to promote error-free normal operations and quick, accurate diagnosis of off-normal conditions.

d. Meet applicable NRC requirements related to engineered safety system design and analysis of plant and engineered safety system responses to regulatory transients and accidents.

e. Evaluate the mean annual core damage frequency (CDF) and large release frequency (LRF) for the APR1400 design using a probabilistic risk assessment (PRA). The design target for CDF is 1 x 10-5 events per reactor year, and the design target for LRF is 1 x 10-6 events per reactor year. These targets include an 1.2-1 Rev. 2

APR1400 DCD TIER 2 assessment of internal and external events, excluding seismic events, sabotage, and other external events, and an assessment of shutdown events.

f. Provide a large, rugged reactor containment building and associated containment systems for heat removal and retention of fission products for design basis events (DBEs) and beyond DBEs (BDBEs). Containment design pressure is based on the most limiting loss of coolant or steam line break accident.

g. Provide containment system components for which a change of state is necessary (e.g., containment isolation valves) that are redundant and sufficiently independent from the systems whose failure could lead to core damage in order to provide reasonable assurance of an intact containment and avoid significant vulnerability to common cause failure.

h. Design the containment systems so the applicable exposure limits are met assuming a reactor containment building design leak rate of no less than 0.1 volume percent per day.

i. Provide at least two separate and independent ac power connections to the grid to decrease the likelihood of a loss of offsite power (LOOP).

j. Reduce the risk of a station blackout (SBO) by providing an independent, safety-related, onsite ac power generation source for each division and by providing a non-safety-related, alternate ac (AAC) onsite power source.

k. Provide adequate severe accident protection through conservatisms inherent in the design and additional plant features that limit direct containment heating, provide reasonable assurance of core debris coolability, and avoid detonable concentrations of hydrogen.

1.2.1.1.2 Performance Design Objectives The performance design objectives of the APR1400 are as follows:

a. Provide a lifespan of 60 years without the need for an extended refurbishment outage.

b. Provide the capability of operating on a fuel cycle, from post-refueling startup to the subsequent post-refueling startup, with a refueling interval of 18 months.

1.2-2 Rev. 2

APR1400 DCD TIER 2 1.2.1.2 Fundamental Design Approach The following subsections describe the fundamental design approaches that were used as the basis for the development of a comprehensive set of technical requirements for the APR1400 design.

1.2.1.2.1 Simplification The approach to the APR1400 design emphasizes simplicity in all aspects of the plant design, construction, and operation. Simplicity is accomplished by pursuing simplification opportunities with high priority and placing greater importance on simplification in design decisions than has traditionally been done.

The APR1400 simplification approaches include the following:

a. Use a minimum number of systems, valves, pumps, instruments, and other types of mechanical and electrical equipment that are consistent with essential functional requirements.

b. Provide a human-system interface that simplifies plant operation and reflects operator needs and capabilities.

c. Provide system and component designs that provide reasonable assurance that the final plant design minimizes demands on the operator during normal operation as well as transient and emergency conditions.

d. Design equipment and arrangements that simplify and facilitate maintenance.

e. Provide protective logic and actuation systems that are more simplified than those in existing plants.

f. Use standardized components to facilitate operations and maintenance.

g. Design for ease and simplification of construction.

1.2.1.2.2 Design Margin The APR1400 design approach includes a consideration of the proper margin that is needed to provide reasonable assurance of plant safety and operability, as follows:

1.2-3 Rev. 2

APR1400 DCD TIER 2

a. Designed capability to accommodate transients without causing initiation of engineered safety systems

b. Ample operator time to assess and deal with upset conditions with minimum potential for damage

c. Enhancement of system and component reliability and minimization of the potential of exceeding limiting conditions for operation (LCO) limits that could cause derating or shutdown 1.2.1.2.3 Safety The APR1400 safety design approach is that there will be excellence in safety to provide reasonable assurance of safety for the general public and personnel. The primary emphasis is on accident prevention, which includes accident resistance and core damage prevention. Emphasis is also placed on mitigation of the consequence of potential accidents so that a balanced approach to safety is achieved.

This design approach of excellence in safety is implemented through an integrated approach that includes three overlapping levels of safety protection-accident resistance, core damage prevention, and mitigation-and therefore, uses a deterministic analysis framework complemented by PRA.

1.2.1.2.4 Proven Technology The APR1400 design approach uses successful, proven technology throughout the plant, including the design of systems and components, maintainability and operability features, and construction techniques.

1.2.2 Principal Site Characteristics The APR1400 is a standard nuclear power plant design that can be constructed on a site with the parameters that are described in Chapter 2. These parameters are the basis for design certification. The site interface parameters presented in Chapter 2 are conservative enough to envelop most potential sites in the United States.

1.2-4 Rev. 2

APR1400 DCD TIER 2 1.2.3 Nuclear Steam Supply System Summary The scope of the APR1400 design covers an essentially complete nuclear power plant that includes all structures, systems, and components (SSCs) that can significantly affect safe operation. The primary design characteristics are summarized in the following subsections. The seismic category, safety classification, and quality assurance requirements of SSCs are listed in Table 3.2-1.

1.2.3.1 Reactor 1.2.3.1.1 Reactor Core The reactor core is fueled by uranium dioxide pellets enclosed in fuel rods. The fuel rods are fabricated into assemblies with nozzles that limit axial motion and grids that limit lateral motion of the fuel rods. The control element assemblies (CEAs) consist of boron carbide (B4C) or Inconel absorber rods that are guided by tubes located within the fuel assembly. The core consists of 241 fuel assemblies that are typically loaded in the first fuel cycles with different U-235 enrichments. The NSSS-rated thermal output is 4,000 MWt with a core thermal output of 3,983 MWt. The reactor core is described in Sections 4.2, 4.3, and 4.4.

1.2.3.1.2 Reactor Internals The reactor internals include the core support barrel, lower support structure and in-core instrumentation nozzle assembly, core shroud, and upper guide structure assembly. The core support barrel is a right circular cylinder supported by a ring flange from a ledge on the reactor vessel. The lower support structure transmits the entire weight of the core to the core support barrel by means of a beam structure. Snubbers are provided at the lower end of the core support barrel to restrict lateral and torsional movement. The core shroud surrounds the core and minimizes the amount of bypass flow. The upper guide structure provides a flow shroud for the CEAs and limits upward motion of the fuel assemblies.

The principal design bases for the reactor internals are to provide vertical supports and horizontal restraints during all normal operating, upset, emergency, and faulted conditions.

The core is supported and restrained during normal operation and postulated accidents to provide reasonable assurance that coolant can be supplied to the coolant channels for heat removal. The reactor internals are described in further detail in Sections 3.9 and 4.5.

1.2-5 Rev. 2

APR1400 DCD TIER 2 1.2.3.2 Reactor Coolant System and Connecting System 1.2.3.2.1 Reactor Coolant System The reactor coolant system (RCS) is arranged as two closed loops connected in parallel to the reactor vessel. Each loop consists of one outlet hot leg, one steam generator (SG), two cold legs, and two reactor coolant pumps (RCPs). A pressurizer (PZR) is connected to one of the RCS loops.

The RCS operates at a nominal pressure of 158.2 kg/cm2A (2,250 psia). The reactor coolant enters the reactor vessel, flows downward between the reactor vessel shell and core barrel, flows up through the core, leaves the reactor vessel, and flows through the tube side of the two SGs where heat is transferred to the secondary system. RCPs return the reactor coolant to the reactor vessel.

Two SGs, using heat generated by the reactor core, produce steam for driving the plant turbine generator (T/G). Each SG is a vertical U-tube heat exchanger with an integral economizer that operates with the reactor coolant on the tube side and secondary coolant on the shellside. Each unit is designed to transfer heat from the RCS to the secondary system to produce saturated steam when provided with the proper feedwater (FW) input.

Moisture separators and steam dryers on the shellside of the SG limit the moisture content of the steam during normal operation. An integral flow restrictor is included in each SG nozzle to restrict flow in the event of a steam line break.

The SG incorporates high-performance steam dryers to limit the moisture content to below 0.25 percent in the steam flow. The heat transfer tubes are made of Alloy 690 TT, which is resistant to stress corrosion cracking in high-temperature conditions. The secondary FW inventory is increased to extend the dry-out time to enhance the NSSS capability to tolerate upset conditions and improve operational flexibility. The heat transfer area is large enough to allow the NSSS to maintain a rated output even if 10 percent of the tubes are plugged.

The RCS is described in further detail in Chapter 5.

1.2-6 Rev. 2

APR1400 DCD TIER 2 1.2.3.2.2 Reactor Coolant System High Point Vents The high point vent system is a dedicated safety system designed to perform the following functions:

a. A safety-grade means of venting non-condensable gases and steam from the PZR and the reactor vessel closure head.

b. A safety-grade means to depressurize the RCS in the event the PZR spray is unavailable during plant cooldown to cold shutdown.

The reactor coolant gas vent system (RCGVS) is described in further detail in Subsection 5.4.12.

1.2.4 Engineered Safety Features Engineered safety features (ESF) are provided to mitigate the consequences of design basis accidents. These ESFs are designed to localize, control, mitigate, or terminate such accidents in order to hold exposure levels below the limits of 10 CFR 50.34 (Reference 1).

1.2.4.1 Reactor Containment Building General arrangements for the reactor containment building are shown in Subsection 1.2.14.

The APR1400 reactor containment building is a steel-lined prestressed concrete structure that consists of a right circular cylinder with a hemispherical dome on a reinforced concrete basemat. The cylindrical portion of the containment structure is prestressed by a post-tensioning system that consists of horizontal (hoop) and vertical (inverted, U-shaped) tendons. The interior surfaces of the containment shell, dome, and basemat are lined with a carbon steel plate for leak-tightness. A protective layer of concrete (filled slab) covers the portion of the liner over the foundation slab. The containment building provides biological shielding for normal and accident conditions.

The containment building completely encloses the reactor and RCS, and is designed to provide a barrier that is essentially leak-tight to the release of radioactive materials subsequent to postulated accidents. The internal structures and compartment arrangement provide equipment missile protection and biological shielding for maintenance personnel.

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APR1400 DCD TIER 2 The containment building is designed for all credible loading combinations, including normal loads during a LOCA, test loads, and loads due to adverse environmental conditions.

1.2.4.2 Safety Injection System The safety injection system (SIS) is designed to satisfy NRC requirements. The requirements are specified as the licensing design basis for the APR1400 design.

In the unlikely event of a LOCA, the SIS injects borated water into the RCS. The SIS incorporates a four-train safety injection configuration and an in-containment refueling water storage tank (IRWST).

The SIS uses four safety injection (SI) pumps to inject borated water directly into the reactor vessel. In addition, four safety injection tanks (SITs) are provided. The SI pumps are aligned to the IRWST, and realignment for recirculation following a LOCA is not required. The SIT provides cooling to limit core damage and fission product release and reasonable assurance of an adequate shutdown margin. The fluidic device (FD) in the SIT regulates the flow rate into the reactor vessel to improve cooling effectiveness.

The SIS also provides continuous long-term, post-accident cooling of the core by recirculating borated water from the IRWST. Water drawn from the IRWST by the SI pumps and containment spray (CS) pumps is injected into the reactor vessel and containment. The SI water then enters the containment through the primary pipe break.

This water and the CS water return through floor drains and the holdup volume tank (HVT) to the IRWST. During this process, heat is removed from the IRWST water by the CS heat exchanger.

The SIS is capable of providing an alternate means of decay heat removal for the events beyond the licensing design basis in which the SGs are not available. Decay heat removal is accomplished by feeding and bleeding the RCS, using the SIS to feed and the pressurizer pilot operated safety relief valve (POSRV) to bleed, and by cooling the IRWST water using the shutdown cooling system (SCS).

The SIS and the IRWST are described in further detail in Sections 6.3 and 6.8, respectively.

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APR1400 DCD TIER 2 1.2.4.3 Auxiliary Feedwater System The auxiliary feedwater system (AFWS) provides feedwater from the auxiliary feedwater storage tanks (AFWSTs) to the SGs for heat removal when the FW system is inoperable for a transient or postulated accident condition.

The AFWS consists of two 100 percent capacity motor-driven pumps, two 100 percent capacity turbine driven pumps, two 100 percent AFWSTs, valves, two cavitating flow-limiting venturis, and instrumentation. Each pump takes suction from the respective AFWST and has a respective discharge header. Each pump discharge header contains a pump discharge check valve, flow modulating valve, auxiliary feedwater (AFW) isolation valve, and SG isolation check valve.

The AFWS components are located in seismic Category I structures, which protect the components from external environmental hazards such as earthquakes, tornados, floods, and external missiles. Each train of the AFWS is physically separated from the others within these structures.

One motor-driven pump train and one turbine-driven pump train are configured into one mechanical division and joined inside the containment to feed their respective SG through a common AFW header, which connects to the steam generator downcomer feedwater line.

Each common AFW header contains a cavitating venturi to restrict the maximum AFW flow rate to each SG.

The AFWS is designed to be manually or automatically actuated by an auxiliary feedwater actuation signal (AFAS). At the low water level setpoint of the SG, the AFAS associated with that SG is designed to actuate the AFWS.

For design basis considerations, sufficient feedwater can be provided at the required temperature and pressure even if a secondary pipe break event occurs. Because the AFWS is the only safety-related source of makeup water to the SGs for heat removal when the FW system is inoperable for a transient or postulated accident condition, it has been designed with redundancy, diversity, and separation to provide reasonable assurance of its ability to function.

The AFWS is described further in Subsection 10.4.9.

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APR1400 DCD TIER 2 1.2.4.4 Containment Spray System The containment spray system (CSS) is designed to maintain containment pressure and temperature within the design limits in the unlikely design basis accidents (DBAs) that result in mass-energy releases to the containment atmosphere. The CSS also provides a containment air cleanup function to reduce the concentration of fission products in the containment atmosphere after an accident.

The CSS consists of two independent trains, each containing a CS pump, a CS heat exchanger, a CS pump mini-flow heat exchanger, spray headers, spray rings and nozzles, and associated valves, piping, and instrumentation.

The CS pumps are automatically actuated on receipt of a safety injection actuation signal (SIAS) or a containment spray actuation signal (CSAS). Upon a CSAS, the containment spray isolation valves open and the CS flow starts. The essential components of the CSS are powered from the emergency power sources to provide assurance of the reliability of the safety function for a loss of offsite power. The suction isolation valve from the IRWST is locked open during power operation. Two CS pumps supply water through two CS heat exchangers to the spray headers in the upper region of the containment. Spray headers are used to provide a relatively uniform distribution of spray over the cross-sectional area of the containment. The IRWST provides a continuous suction source for the CS pumps, thus eliminating the switchover from the IRWST to the containment recirculation sump for conventional PWR plants.

The CS pumps can be manually aligned and used as residual heat removal pumps during SCS operation. Likewise, the SC pumps can be manually aligned to perform the containment spray function.

The CS pumps can also be used as a backup to the SC pumps to provide cooling of the IRWST water during post-accident feed and bleed operations when the SGs are not available to cool the RCS.

The CSS is further discussed in Subsections 6.2.2 and 6.5.2.

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APR1400 DCD TIER 2 1.2.4.5 Containment Isolation System The containment isolation system (CIS) provides means of isolating fluid systems that pass through the containment penetrations to confine the release of any radioactivity from the containment following a postulated DBA.

In conformance to 10 CFR Part 50, Appendix A, General Design Criterion (GDC) 54 (Reference 2), the piping systems and related components penetrating the containment are provided with leak detection, isolation, and containment capabilities with redundancy, reliability, and performance capabilities that reflect the safety-related importance of isolating these fluid systems.

Isolation design is achieved by applying acceptable common criteria to penetrations in many different fluid systems and by using containment pressure to provide a containment isolation actuation signal (CIAS) to actuate appropriate valves.

The CIS is described further in Subsection 6.2.4.

1.2.4.6 Engineered Safety Features Filter Systems ESF filters are provided for the systems that are required to perform safety-related functions subsequent to a DBA, as follows:

a. Control room emergency makeup air cleaning system The system is part of the control room heating, ventilation, and air conditioning (HVAC) system and is used to clean up the makeup air that has potential to carry radioactive iodine and particulates following a DBA.

The system is normally shut down and starts automatically in response to any one of the following signals:

1) SIAS

2) Control room emergency ventilation actuation signal (CREVAS)

3) Remote manual activation from the main control room (MCR)

b. Auxiliary building controlled area emergency exhaust system 1.2-11 Rev. 2

APR1400 DCD TIER 2 The system is part of the auxiliary building controlled area HVAC system and is used to filter radioactive elemental iodines and particulates in the exhaust air from the safety-related mechanical equipment rooms, which are cooled by safety-related cubicle coolers after a DBA.

The system is normally shut down and starts automatically in response to any one of the following signals:

1) SIAS

2) Remote manual activation from the MCR

c. Fuel handling area emergency exhaust system The system is part of the fuel handling area HVAC system and is used to reduce the radioactive elemental iodines and particulates in the exhaust air from the fuel handling area following a fuel handling accident.

The system is normally shut down and starts automatically in response to one of the following signals:

1) High radiation signal from the radiation monitor located in the common discharge duct of the fuel handling area exhaust air cleaning units (ACUs)

2) Fuel handling area emergency ventilation action signal (FHEVAS)

3) Remote manual activation from the MCR ESF filter systems are described further in Subsection 6.5.1.

1.2.5 Instrumentation and Control 1.2.5.1 Reactor Trip System The reactor trip system (RTS) is a safety system that initiates reactor trips. The RTS consists of four channels of sensors, auxiliary process cabinet-safety (APC-S), ex-core neutron flux monitoring system (ENFMS), core protection calculator system (CPCS), the reactor protection system (RPS) portion of the plant protection system (PPS), and reactor trip switchgear system (RTSS).

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APR1400 DCD TIER 2 Four independent channels of the RPS monitor the selected plant parameters. The RPS logic is designed to initiate protective action whenever the signals of any two channels of a given parameter reach the setpoint. If this occurs, the power supplied to the control element drive mechanisms (CEDMs) is interrupted through the RTSS. The CEDMs release the CEAs, which drop into the core to shut down the reactor.

1.2.5.2 Engineered Safety Features System The engineered safety features (ESF) system consists of four channels of sensors, APC-S, the engineered safety features actuation system (ESFAS) portion of PPS, and the engineered safety features - component control system (ESF-CCS).

The ESF-CCS accepts ESFAS initiation signals from the ESFAS portion of the PPS and radiation monitoring system (RMS). The ESF actuation logic is used to activate ESF system components of the plant. Emergency diesel generator (EDG) loading sequencer logic is also included in the ESF-CCS. The component control logic in the ESF-CCS is described in Subsection 1.2.5.3. The ESF actuation logic has a selective 2-out-of-4 coincidence logic for the NSSS ESFAS or 1-out-of-2 logic for the BOP ESFAS so that no single failure can preclude the system from providing the safety function. The ESF actuation signal actuates ESF system components through the ESF-CCS.

1.2.5.3 Component Control System The component control system (CCS) is designed to provide control of plant process components and to acquire data on the process components. The CCS provides discrete and continuous control of plant components.

The CCS consists of the ESF-CCS and process-CCS (P-CCS) assemblies to provide control for the different divisions of safety equipment, as well as non-safety equipment. Although the safety and non-safety CCS assemblies perform different plant control functions, they use diverse software and software-dependent electronic components.

1.2.5.4 Diverse Protection System The diverse protection system (DPS) augments the plant protection function by initiating a reactor trip signal, turbine trip signal, AFAS, and SIAS that are separate and diverse from the PPS.

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APR1400 DCD TIER 2 The DPS is provided to address the design requirements of 10 CFR 50.62 (Reference 3) and the Staff Requirements Memorandum (SRM) regarding SECY-93-087, II.Q (Reference 4).

The DPS equipment provides a simple and diverse mechanism to significantly decrease risk from anticipated transient without scram (ATWS) events and assist the mitigation of the effects of a postulated common-cause failure (CCF) of the digital computer logic within the PPS and ESF-CCS.

The DPS initiates a reactor trip when the PZR or containment pressure exceeds a predetermined value. For implementation of the reactor trip function, the DPS circuitry is diverse from the PPS, from sensor output to interruption of power to control rods. The DPS design uses a 2-out-of-4 logic to open trip circuit breakers of the reactor trip switchgear system (RTSS).

The DPS initiates the AFAS when the level in either SG decreases below a predetermined value and initiates the SIAS when the PZR pressure decreases below a predetermined value.

From sensor output to, but not including, the final actuation device, the DPS circuitry for the AFAS and SIAS is independent and diverse from the circuitry of the PPS and ESF-CCS.

1.2.5.5 Reactor Control Systems The startup, operation, and shutdown of the reactor are accomplished through integrated control system actions. These control systems regulate reactor power and respond to plant transients to maintain the NSSS within its normal operating conditions. Reactor control functions are performed by the power control system (PCS) and NSSS process control system (NPCS) of the P-CCS, as described in Section 7.7. The PCS performs digital rod control system (DRCS), reactor power cutback system (RPCS), and required response spectra functions to adjust the reactor power response to turbine load demand. The NPCS performs steam bypass control system (SBCS), feedwater control system (FWCS), and PZR control functions.

Reactor power control is normally accomplished by the automatic movement of CEAs in response to a change in reactor coolant temperature, with manual control that is capable of overriding the automatic signal at any time. If the reactor coolant temperature is different from a programmed value, the CEAs are adjusted until the difference is within the prescribed control band. Regulation of the reactor coolant temperature, in accordance with this process, maintains the secondary steam pressure within operating limits and matches reactor power to load demand.

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APR1400 DCD TIER 2 The reactor is controlled by a combination of CEA motion and dissolved boric acid in the reactor coolant. Boric acid is used for reactivity changes associated with large but gradual changes in water temperature, xenon concentration, and fuel burnup. The addition of boric acid also provides an increased shutdown margin during the initial fuel loading and subsequent refuelings. The boric acid solution is prepared and stored at a temperature that prevents precipitation.

CEA movement provides changes in reactivity for shutdown or power changes. The CEAs are moved by CEDMs mounted on the reactor vessel head. The CEDMs are designed to permit rapid insertion of the CEAs into the reactor core by gravity. CEA motion can be initiated manually or automatically.

The pressure in the RCS is controlled by regulating the temperature of the coolant in the PZR where steam and water are maintained in thermal equilibrium. Steam is formed by the PZR heaters or condensed by the PZR spray to reduce variations caused by expansion and contraction of the reactor coolant because of temperature changes.

The SBCS is used to dump steam in case of a large mismatch between the power being produced by the reactor and the power being used by the turbine. Dumping steam allows the reactor to remain at power instead of tripping. The water level in each SG is maintained by the FWCS. The RPCS is used to drop selected CEAs into the core to reduce reactor power rapidly during the large loss of load or failure of 2-out-of-3 operating FW pumps. Dropping selected CEAs allows the SBCS and the FWCS to maintain the NSSS in a stable condition without a reactor trip and without lifting any safety valves during the transients after the loss of load.

1.2.5.6 Nuclear Instrumentation The nuclear instrumentation includes ex-core and in-core neutron flux detectors and associated signal processing equipment. Eight channels of ex-core instrumentation monitor the power. Two startup channels are provided for startup, two control channels are provided for power control, and four safety channels are provided for protection.

The startup channels are used to monitor the power that is used during the initial reactor startup, extended shutdown periods, startup after extended periods of shutdown, and after refueling operations. The control channels are used to control the reactor power during power operation. The safety channels are used to provide inputs to the variable overpower, 1.2-15 Rev. 2

APR1400 DCD TIER 2 high logarithmic power, low departure from nucleate boiling ratio (DNBR), and high local power density (LPD) trips in the RPS.

The in-core nuclear instrumentation consists of fixed in-core nuclear instrumentation detectors distributed throughout the core. The instrumentation is used to monitor the power distribution in the core and evaluate fuel burnup in each fuel assembly and thermal margins in the core.

1.2.5.7 Process Monitoring Systems Temperature, pressure, flow, and liquid level are monitored as required to keep operating personnel informed of plant operating conditions. Protection channels indicate the various parameters used for protective action and provide trip and pre-trip alarms from the RPS.

Plant liquid and gaseous effluents are monitored to provide reasonable assurance that they are maintained within applicable radioactivity limits. Additional information is provided in Section 11.5.

1.2.6 Human-System Interface System 1.2.6.1 Main Control Room The MCR is provided with a redundant, compact, workstation-type human-system interface (HSI); large display panel (LDP); safety console; voice communication equipment; and other equipment that is necessary for safe and reliable plant operation.

Qualified indication and alarm displays are provided to permit normal and accident plant operations in the unlikely event that the information processing system (IPS) becomes unavailable.

The arrangements and layouts for all controls and displays in the MCR are designed, verified, and validated in accordance with human factors design guidelines and the requirements in the APR1400 human factors engineering program plan, which is described in Section 18.1. The layout of an MCR is shown in Figure 7.7-13.

An operator workstation, including sufficient desk space, is provided to support the plant monitoring and daily operational needs for each operator.

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APR1400 DCD TIER 2 1.2.6.2 Remote Shutdown Room The RSR design includes the remote shutdown console (RSC), which is similar to the reactor operator (RO) workstation in the MCR, and the shutdown overview display panel (SODP) to achieve cold shutdown (Mode 5 plant conditions) when operators are to evacuate the MCR. The layout of the RSR is shown in Figure 7.4-4.

For a safe shutdown from the RSR, controls and indications are available through information on flat panel displays (FPDs) and soft controls on the RSC. The SODP provides information that the operator uses during plant shutdown operation.

For consistency, the information displays and soft controls on the RSC are the same as in the MCR.

1.2.6.3 Qualified Indication and Alarm System The qualified indication and alarm system (QIAS) is composed of the qualified indication and alarm system - P (QIAS-P) and qualified indication and alarm system - non-safety (QIAS-N).

The QIAS-P provides a continuous and dedicated display of NRC RG 1.97 (Reference 5)

Type B and C parameters for accident monitoring.

The QIAS-N receives analog and digital data from both safety and non-safety systems, analyzes the data, and relays the results of the analysis to the operator via the QIAS-N FPD and mini-LDP in the MCR, and SODP in the RSR. The system interfaces with the IPS to integrate alarm and process information.

Additional information is provided in Section 7.5.

1.2.6.4 Information Processing System The IPS is a fault-tolerant, multi-processor, computer-based system that provides plant data and status information to the operating staff. The IPS monitors the NSSS and balance-of-plant (BOP) steam and electrical production processes. The IPS provides plant operating staff the ability to obtain detailed process data via FPDs and LDPs.

The major functions performed by the IPS include plant-wide data acquisition through dedicated data links to plant systems, validation of sensed parameters, execution of 1.2-17 Rev. 2

APR1400 DCD TIER 2 application programs and performance calculations, monitoring of general plant status and plant safety status, generation of logs and reports, determination of alarm conditions, recording of the sequence of events, and generation of a post-trip review.

FPD and LDP formats incorporate human factors engineering design principles that permit quick operator recognition of information that is necessary to allow the operator to monitor, control, and diagnose plant conditions.

The IPS is designed to provide the plant operating staff with reliable, complete, and timely information for the safe and efficient operation of the plant. The IPS is designed to tolerate the loss of any single major system component without total loss of functionality.

The design includes automatic fail-over and sufficient redundant peripherals to minimize the effects of an IPS component failure during plant operations.

The IPS is described further in Section 7.7.

1.2.7 Electrical System Offsite and onsite power systems are provided to supply electrical power to unit auxiliaries that are necessary during normal operation and the RPS and ESF that are necessary in abnormal and accident conditions.

The offsite power system consists of transmission lines, transmission line towers, switchyard components and a control system, switchyard battery systems, transmission tie lines, main generator, generator circuit breaker (GCB), main transformer, unit auxiliary transformers (UATs), and standby auxiliary transformers (SATs).

Under normal operating conditions, the main generator supplies power through an isolated phase bus and the GCB to the main transformer and UATs. The UATs are connected to the isolated phase bus between the GCB and main transformer. Additional information on the offsite power system is provided in Section 8.2.

The onsite power system for the unit auxiliaries consists of four EDGs, an alternate alternating current (AAC) gas turbine generator (GTG), and two onsite power distribution systems (a Class 1E system and a non-Class 1E system). The onsite power distribution system is connected to the site-specific switchyard via two separate and independent transmission tie circuits. One circuit is connected to the switchyard through the main transformer and UATs, and the other circuit is connected to the switchyard via the SATs.

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APR1400 DCD TIER 2 During normal operation, onsite power is supplied from the main generator through the UATs. During startup and shutdown, the GCB is open, and the onsite power is supplied from the transmission system through the main transformer and UATs.

The onsite power system is described further in Section 8.3. A description of the AAC generator system is provided in Section 8.4.

1.2.8 Steam and Power Conversion System The function of the steam and power conversion system is to convert heat energy generated by the nuclear reactor into electrical energy. The heat energy produces steam in two SGs capable of driving a turbine-generator (T/G) unit.

The steam and power conversion system consists of the T/G, main steam system (MSS),

condensate and FW system, and other support systems. The steam and power conversion system uses a condensing cycle with regenerative FW heating.

The steam generated in the two SGs is supplied to the high-pressure turbine by the MSS.

The steam is expanded through the high-pressure turbine, passes through the two moisture separator reheaters (MSRs), and then flows to the three low-pressure turbines.

The exhaust steam from the low-pressure turbines is condensed in a conventional surface-type condenser. The condenser removes air and other non-condensable gases from the condensate and transfers heat to the circulating water system.

The condensate from the steam is returned to the SGs through the condensate and FW system. The condensate from the condenser hotwell is transferred through the low-pressure (LP) heaters to the deaerator storage tank by the condensate pumps.

The FW booster pumps take suction from the deaerator storage tank and discharge to the FW pumps. FW is discharged from the FW pumps, passes through two trains of high-pressure FW heaters, and is delivered to the SGs.

The steam and power conversion system is described further in Chapter 10.

1.2.8.1 Turbine Generator The T/G converts the thermal energy of the steam produced in the SGs into mechanical shaft power and then into electrical energy.

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APR1400 DCD TIER 2 The T/G consists of a double-flow, high-pressure turbine and three double-flow, low-pressure turbines driving a direct-coupled generator and two external MSRs.

The flow of main steam is directed from the SGs to the high-pressure turbine through main stop and control valves. After expanding through the high-pressure turbine, exhaust steam passes through the MSRs. Extraction from the high-pressure turbine and main steam from the equalization header is supplied to the first and second stage of reheater tube bundles in each reheater. The hot reheat steam is admitted to the low-pressure turbines through combined intermediate valves and expands through the low-pressure turbines to the main condensers.

The T/G control system is designed to be compatible with the plant control system for reactor operation. The T/G is designed to accept a sudden loss of full load or LOOP without exceeding design overspeed.

The T/G is described further in Section 10.2.

1.2.8.2 Main Steam System The MSS delivers steam generated in the SGs to the high-pressure turbine where the thermal energy of the steam is converted to mechanical energy to drive the main T/G. The MSS also provides steam to the FW pump turbines, AFWP turbines, second-stage reheater of the MSRs, turbine steam seal system, auxiliary steam system, and process sampling system.

The major components of the MSS are the main steam piping, main steam isolation valves (MSIVs), main steam isolation valve bypass valves (MSIVBVs), main steam safety valves (MSSVs), main steam atmospheric dump valves (MSADVs), turbine bypass valves (TBVs),

and AFWP turbine steam supply valves and warmup valves.

An MSIV is installed on each of the main steam lines downstream of the MSSVs, outside the reactor containment building. The MSIVs are provided to isolate the SGs upon receipt of a main steam isolation signal (MSIS). MSIVs are remote-operated and fail-closed valves with a hydraulic actuator.

Overpressure protection for the secondary side of the SGs is provided by spring-loaded MSSVs. Modulation of the TBVs would normally prevent the safety valves from opening.

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APR1400 DCD TIER 2 Following the load rejection of any magnitude from full load to house load, including a turbine trip from 100 percent power, the TBS controls main steam pressure automatically by the SBCS.

During a turbine or reactor trip, the TBS dissipates heat from the reactor coolant system to the condensers. The system has the capability of relieving 55 percent of full load main steam flow to the main condenser.

The MSS is provided with MSADVs to remove reactor decay heat during hot standby and emergency cooldown in conjunction with AFWS.

The MSS is described in further detail in Section 10.3.

1.2.8.3 Condensate and Feedwater System The condensate and feedwater system delivers feedwater from the condenser to the SG.

The entire condensate system is non-safety-related. The portions of the feedwater system that are required to mitigate the consequences of an accident and allow safe shutdown of the reactor are safety-related.

The condensate and feedwater system is described further in Subsection 10.4.7.

1.2.9 Heating, Ventilation, and Air Conditioning System The HVAC systems for all plant buildings are designed for personnel comfort and equipment operation. In addition, the following systems are provided with the protection features described as follows:

a. The control room HVAC system is designed to maintain the environment in the control room envelope and limit the radiation exposure of personnel in the control room during all plant operation conditions. The system maintains positive pressure to provide habitability and prevent uncontrolled incoming air leakage.

b. The fuel handling area HVAC system is a once-through ventilation system designed to limit the radiation release following a fuel-handling accident to meet 10 CFR 50.34 guidelines. This system maintains the area under negative pressure and airflow from less-contaminated to more-contaminated areas.

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APR1400 DCD TIER 2

c. The compound building controlled-area HVAC system is a once-through ventilation system with filtered exhausts. This system maintains negative building pressure and airflow from less-contaminated to more-contaminated areas.

d. The auxiliary building controlled-area HVAC system is a once-through ventilation system designed to filter post-accident contaminated leakages before exiting to meet 10 CFR 50.34 guidelines. This system maintains the building under negative pressure and airflow from less-contaminated to more-contaminated areas.

e. The containment purge system is provided with post-accident containment isolation features and filtration units for air cleanup during normal and refueling operations. This system limits the radiation release to meet 10 CFR 50.34 guidelines in case of a fuel handling accident inside the containment.

Other HVAC systems are described in Section 9.4.

1.2.10 Fuel Handling and Storage 1.2.10.1 Fuel Handling Fuel handling equipment provides for the safe handling of fuel assemblies and CEAs under specified conditions and for the required assembly, disassembly, and storage of the integrated head assembly and reactor internals during refueling.

The major components of the system are the refueling machine, the CEA change platform, the fuel transfer system, the spent fuel handling machine, the new fuel elevator, and the CEA elevator. The fuel handling equipment is provided to transfer new and spent fuel assembly between the fuel storage facility, the containment building, and the fuel shipping and receiving areas during initial core loading and refueling operations. A fuel assembly is inserted into or removed from the core using the refueling machine. During normal operations, irradiated fuel assemblies and CEAs are maintained in a water environment.

The principal design criteria specify the following:

a. Fuel is inserted, removed, and transported in a safe manner.

b. Subcriticality is maintained during all operations.

Fuel handling is further discussed in Section 9.1.

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APR1400 DCD TIER 2 1.2.10.2 Fuel Storage The new fuel storage facility provides onsite storage capacity of 112 new fuel assemblies.

This capacity, which represents 46 percent of the fuel assemblies in the core, envelops a reload batch based on a refueling cycle of 18 months. The spent fuel storage is divided into two regions of the fuel. The fresh or partially burnt fuel assemblies are stored in Region I, which has storage capacity for one full core, one refueling batch, and five damaged fuel assemblies. The Region I storage area is designed to accommodate fuel assemblies with an initial enrichment up to 5 weight percent U-235. Region II has a storage capacity of spent fuel assemblies generated during a plant operation of 20 years.

The maximum initial enrichment of 5 weight percent U-235 and the minimum burnup are applied to the Region II design.

Criticality and safety analyses are addressed in Subsection 9.1.1. The new fuel and spent fuel storage facilities are addressed in Subsection 9.1.2.

1.2.11 Cooling Water Systems 1.2.11.1 Circulating Water System The circulating water system provides cooling water for the condensers and T/G building closed cooling water heat exchangers and rejects heat to the normal plant heat sink. The normal plant heat sink is site-specific, but a mechanical draft cooling tower is used as a preferable heat sink.

The circulating water system is described further in Subsection 10.4.5.

1.2.11.2 Essential Service Water System The essential service water system (ESWS) is an open system that takes suction from the ultimate heat sink (UHS) and provides cooling water to remove heat released from plant SSCs. The ESWS returns the heated water to the UHS. The ESWS cools the component cooling water system (CCWS), which in turn cools essential and non-essential reactor auxiliary loads.

The ESWS consists of two independent, redundant, safety-related divisions. Each division consists of two ESW pumps, three CCW heat exchangers, three ESW debris filters, and associated piping, valves, controls and instrumentation.

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APR1400 DCD TIER 2 During normal operation, one ESW pump and two CCW heat exchangers per division are in service.

During plant shutdown operations, two ESW pumps and three CCW heat exchangers in each division operate to remove heat from the components required for plant shutdown operation.

During plant abnormal operation, one ESW pump and two CCW heat exchangers in a single division operate to remove heat from the essential components required for safe shutdown or mitigation of plant abnormal conditions.

In the event of a LOOP, each division of the ESWS is automatically powered from the EDGs in accordance with emergency load sequencing.

The ESWS is described further in Subsection 9.2.1.

1.2.11.3 Component Cooling Water System The CCWS is a closed-loop cooling water system that, in conjunction with the ESWS and the UHS, removes heat generated from essential and non-essential plant components connected to the CCWS. Heat transferred by these components to the CCWS is rejected to the ESWS via the CCW heat exchangers.

The CCWS consists of two independent, redundant closed loop divisions. Each division consists of two CCW pumps, three CCW heat exchangers, a CCW surge tank, a CCW chemical addition tank, a CCW makeup pump, and associated piping, valves, and instruments.

During normal power operation, one CCW pump and two CCW heat exchangers in each division are in service to supply cooling water to safety-related and non-safety-related components in the division required for normal power operation.

During the plant shutdown operation, two CCW pumps and three CCW heat exchangers in each division operate to supply cooling water to the components required for plant shutdown operation.

During abnormal plant operations, one CCW pump and two CCW heat exchangers in a single division operate to supply cooling water to the essential components required for the safe shutdown of the plant or mitigation of the abnormal condition.

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APR1400 DCD TIER 2 In the event of a LOOP, each division of the CCWS is automatically powered from the EDGs in accordance with emergency load sequencing.

The CCWS is described further in Subsection 9.2.2.

1.2.11.4 Chilled Water System The chilled water system is designed to provide and distribute a sufficient quantity of chilled water, through a group of dedicated piping systems, to air handling units (AHUs) and cubicle coolers in specific plant areas. The system is divided into two subsystems: an essential chilled water system (ECWS) that serves safety-related HVAC cooling loads and a plant chilled water system (PCWS) that serves primarily non-safety-related HVAC cooling loads.

The chilled water system is described further in Subsection 9.2.7.

1.2.11.5 Spent Fuel Pool Cooling and Cleanup System The spent fuel pool cooling and cleanup system (SFPCCS) consists of the spent fuel pool (SFP) cooling system and the SFP cleanup system.

The safety-related SFP cooling system consists of two redundant trains that are independent of each other. The SFP cooling system removes decay heat generated by one full core offloaded after 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> following shutdown, plus the spent fuel assemblies accumulated from the previous refueling operations. Spent fuel is placed in the pool during the refueling operation and stored there until shipped offsite. Heat is transferred from the SFP cooling system, through an SFP cooling heat exchanger, to the CCWS. When a cooling train is in operation, water flows from the SFP to the SFP cooling pump suction, is pumped through the hot side of the heat exchanger, and is returned to the SFP. The suction line is located at an elevation above the required minimum water level, while the return line contains an anti-siphon device to prevent gravity drainage of the SFP.

The non-safety-related SFP cleanup system consists of pumps, demineralizers, and filters to maintain SFP water clarity and purity. Fuel transfer canal and refueling pool water is circulated through the same demineralizers and filters. These cleanup loops are sufficient for removing fission products and other contaminants that may be introduced if a leaking fuel assembly is transferred to the SFP.

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APR1400 DCD TIER 2 The demineralizer and filter of the cleanup train are used to clean and purify the SFP water or refueling water while SFP heat removal operations proceed. Connections are provided so that the water may be pumped from either the IRWST or the SFP through a filter and demineralizer and discharged to IRWST or the SFP. To assist further in maintaining SFP optical clarity, the SFP surface is cleaned by a skimmer.

The SFP receives borated makeup water from the boric acid storage tank (BAST) through the chemical and volume control system (CVCS). The seismic Category I backup source is provided from the AFWST via the CCW makeup pumps. The non-seismic Category source of nonborated demineralized water to the SFP is available during normal plant conditions.

The SFPCCS is described further in Subsection 9.1.3.

1.2.12 Auxiliary Systems 1.2.12.1 Shutdown Cooling System The SCS is used to reduce the temperature of the reactor coolant, at a controlled rate, from the hot shutdown operating temperature to the refueling temperature and to maintain the proper reactor coolant temperature during refueling. The system uses SC pumps to circulate the reactor coolant through two SC heat exchangers and return it to the RCS.

The CCWS supplies cooling water for the SC heat exchangers.

The SCS has a design pressure of 63.28 kg/cm2G (900 psig). The system pressure provides for greater operational flexibility and simplifies concerns about system overpressurization.

The SCS is described in further detail in Subsection 5.4.7.

1.2.12.2 Chemical and Volume Control System The CVCS controls the purity, volume, and boric acid content of the reactor coolant. The CVCS is not required for any safe shutdown or accident mitigation function.

The coolant purity level in the RCS is controlled by continuous purification of a bypass stream of reactor coolant. Water removed from the RCS is cooled in the regenerative heat exchanger and letdown heat exchanger.

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APR1400 DCD TIER 2 From there, the coolant flows through a filter and a demineralizer where corrosion and fission products are removed. The coolant is then sprayed into the volume control tank (VCT) and returned by the charging pumps to the regenerative heat exchanger for heating prior to returning to the RCS loops. A portion of the flow downstream of the charging pump is diverted for RCP seal injection. The charging flow is controlled by centrifugal charging pumps and a charging flow control valve on the discharge of the pumps.

The CVCS automatically adjusts the amount of reactor coolant in order to maintain a programmed level in the PZR.

The CVCS controls the boric acid concentration in the coolant by feed and bleed where the purified letdown stream is diverted to a boron recovery subsystem, and either concentrated boric acid or demineralized water is sent to the charging pumps. The diverted coolant stream is processed by ion exchange and degasification and flows to a concentrator. The concentrator bottoms are sent to the BAST for reuse as boric acid solution, and the distillate is passed through an ion exchanger and stored for reuse as demineralized water in the reactor makeup water tank.

Moving accident mitigation and safe-shutdown functions to other dedicated safety systems has permitted simplification of plant systems. Although not a safety-related system, the CVCS could provide makeup and depressurization capabilities.

A CVCS is described in further detail in Subsection 9.3.4.

1.2.12.3 Primary Sampling System The primary sampling system is designed to collect and deliver representative samples for inline and laboratory analyses. Typical results of the analyses include reactor coolant boron and chloride concentrations, fission product radioactivity level, radionuclide gamma-spectrum, dissolved gas concentrations, fission gas content, conductivity, pH, corrosion product concentration, and chemical additive concentration. The analysis results are used in regulating boron concentration, evaluating fuel element integrity and demineralizer performance, maintaining acceptable hydrogen levels, detecting radioactive material leakage, and regulating additions of corrosion-controlling chemicals to the systems.

The system consists of sampling lines, a normal primary sample sink, a normal primary sample cooler rack, post-accident primary sample cooler rack, post-accident primary 1.2-27 Rev. 2

APR1400 DCD TIER 2 sample sink, normal/post-accident primary sample control panels, primary off-gas hydrogen/oxygen analyzer, analysis equipment, and associated valves and instrumentation.

The system permits sampling during reactor operation, cooldown, and post-accident modes without requiring access to containment. Remote samples of fluids can be taken from high radiation areas without requiring access to these areas. Local sampling points are provided at various locations throughout the plant. Samples from the containment flow through containment isolation valves to the post-accident primary sample room in the auxiliary building or the normal primary sample room in the compound building.

High-temperature sample lines also contain sample coolers in the normal and post-accident primary sample cooler racks.

The primary sampling system is described further in Subsection 9.3.2.

1.2.12.4 Condensate Polishing System The condensate polishing system (CPS) is designed to remove dissolved and suspended impurities that could cause corrosion damage to secondary system equipment. Condensate polishing demineralizers are also used to remove impurities that enter the system as a result of a condenser circulating water tube leak.

The condensate polishing system is described further in Subsection 10.4.6.

1.2.12.5 Steam Generator Blowdown System The SG blowdown system (SGBS) is designed to assist in maintaining the chemical characteristics of the secondary side water within permissible limits during normal operation and anticipated operational occurrences (AOOs) such as a main condenser tube leak or SG primary-to-secondary tube leakage. The SGBS is also designed to remove impurities concentrated in SGs by continuous blowdown (CBD), periodical high-capacity blowdown (HCBD), and emergency blowdown (EBD).

The SGBS consists of the blowdown subsystem (BDS) and wet layup subsystem (WLS).

The BDS consists of blowdown piping connected to each SG, a blowdown flash tank, a regenerative heat exchanger, two pre-filters, two demineralizers, a post-filter, and control valves. The WLS consists of two recirculation trains (one for each SG) and shares filters and demineralizers with the BDS.

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APR1400 DCD TIER 2 During normal operations, the CBD (0.2% or 1% of the full-power main steam flow) flows from each SG are maintained in order to keep SG the secondary side water chemistry within the specified limits.

The blowdown is directed into a flash tank where the flashed steam is returned to the cycle via the high-pressure FW heaters. The liquid portion flows to a heat exchanger for cooling and is directed through a blowdown filter where a major portion of the suspended solids is removed. After filtration, the blowdown fluid is processed by blowdown demineralizers and returned to the condenser. During long-term shutdown periods, the WLS is used to control water chemistry in the SGs. Following draining or dry layup, the WLS is used to refill the SGs.

The blowdown lines from the SGs are automatically isolated by closing isolation valves in the event of abnormal conditions.

The SGBS is described further in Subsection 10.4.8.

1.2.12.6 Compressed Air and Gas Systems The compressed air and gas systems comprise the compressed air system, the compressed gas system, and the breathing air system. The compressed air and gas systems are non-safety-related with the exception of containment penetration portion.

The instrument air system supplies clean, oil-free, dried air to all air-operated instrumentation and valves. The service air system supplies compressed air for air-operated tools, miscellaneous equipment, and various maintenance purposes.

The compressed gas system comprises the nitrogen subsystems, hydrogen subsystem, and carbon dioxide subsystem.

The breathing air system supplies emergency breathing air for control room personnel.

The compressed air and instrument air systems are described further in Subsection 9.3.1.

1.2.12.7 Equipment and Floor Drainage System The equipment and floor drainage system (EFDS) provides the means by which wastes are appropriately segregated and transported to the liquid waste management system (LWMS) to minimize liquid and gaseous radioactive releases.

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APR1400 DCD TIER 2 The EFDS is described further in Subsection 9.3.3.

1.2.12.8 Fire Protection Program The fire protection program protects SSCs important to safety from the effects of a potential fire. The plant achieves safe shutdown with the assumption that fire will render all equipment in any one fire area inoperable, recognizing that postfire reentry for repairs or operator action will not be possible. The plant also maintains the ability to minimize the potential for radioactive releases to the environment in the event of a fire.

The fire protection program includes administrative controls, emergency lighting, fire barriers, fire detection and suppression systems, fire brigade personnel, and other features provided for fire protection purposes.

The fire protection program is described further in Subsection 9.5.1.

1.2.12.9 Communication Systems The communication systems are designed to provide effective communications between all areas of the plant and the plant site, including all vital areas of the plant. In addition, the communication systems are designed to provide an effective means to communicate to plant personnel and offsite utility and regulatory officials during normal conditions, abnormal, and accident conditions.

The communication systems are described further in Subsection 9.5.2.

1.2.12.10 Lighting System The lighting system is designed to provide adequate and effective illumination throughout the plant and plant site, including all vital areas of the plant.

The normal lighting system is used to provide normal illumination under normal plant operation, maintenance, and test conditions.

Upon loss of the normal lighting system, the emergency lighting system is used to provide acceptable levels of illumination throughout the station and particularly in areas where emergency operations are performed, such as control rooms, fuel handling area, remote shutdown area, and Class 1E switchgear rooms.

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APR1400 DCD TIER 2 The lighting system is described further in Subsection 9.5.3.

1.2.12.11 Emergency Diesel Generator System The EDG system is a safety-related system consisting of four EDGs and their respective support systems such as fuel oil, lube oil, engine cooling water, starting air, and combustion air intake and exhaust systems. Each EDG provides Class 1E power to one of the four independent Class 1E buses during a LOOP. EDGs are normally in standby mode.

Each EDG is designed to attain the rated voltage and frequency within 17 seconds of a loss of voltage, and to be connected to the 4.16 kV Class 1E bus within 19 seconds of a loss of voltage.

Once the EDG reaches rated voltage and speed, the EDG breaker closes and the sequencer generates the proper signal to connect ESF equipment to the Class 1E bus in a programmed time sequence.

The EDG support systems are described further in Subsections 9.5.4 through 9.5.8.

1.2.12.12 Gas Turbine Generator Facility One GTG is used as an AAC source to cope with an SBO. The GTG is independent from the EDGs. The GTG manually starts from a standby condition, accelerates to the required speed, reaches nominal voltage and frequency, and is ready to accept load within 2 minutes of receipt of a start signal in the event of an SBO. The GTG is also designed to start automatically and to be connected manually to non-Class 1E cables in the event of a LOOP.

The major components of the GTG are a combustion turbine, generator, and auxiliary systems such as fuel oil, lube oil, start system, and combustion air intake and exhaust systems.

The GTG support systems are described further in Subsection 9.5.9.

1.2.12.13 Domestic Water and Sanitary System The domestic water and sanitary system provides water for drinking and sanitary purposes.

The sanitary system is designed to receive and treat sewage. This system serves no safety functions and any malfunction has no adverse effect on any safety-related system. The 1.2-31 Rev. 2

APR1400 DCD TIER 2 requirements of 10 CFR Part 50, Appendix A, GDC 60 (Reference 6) are met as related to the design provisions provided to control the release of liquid effluents containing radioactive material from contaminating the domestic water and sanitary system.

The domestic water and sanitary system is described further in Subsection 9.2.4.

1.2.13 Radioactive Waste Management Systems The radioactive waste management systems are designed to control radioactive liquid, gaseous, and solid wastes. The systems consist of three principal systems:

a. Liquid waste management system

b. Gaseous waste management system

c. Solid waste management system The solid, gaseous, and liquid waste management systems are located in the compound building. The design of the radioactive waste management systems provides reasonable assurance that the total offsite dose resulting from radioactive releases is as low as is reasonably achievable (ALARA).

1.2.13.1 Liquid Waste Management System The liquid waste management system (LWMS) is designed to monitor, control, collect, process, handle, store, and dispose of liquid radioactive waste generated during normal plant conditions, including AOOs. The LWMS is divided into the floor drain subsystem, equipment waste subsystem, chemical waste subsystem, and detergent waste subsystem.

The LWMS treats liquid waste using a reverse osmosis (R/O) package system that reduces radioactivity to levels acceptable for release or reuse. The processed liquid radioactive waste is sampled prior to release from monitor tanks. The LWMS is designed to monitor radioactivity levels in the processed liquid waste prior to release.

The LWMS meets the following design requirements:

a. Capability to process floor drain wastes, equipment wastes, chemical wastes, and detergent wastes to meet release radionuclide concentration limits in accordance with 10 CFR Part 20, Appendix B (Reference 7), prior to discharge to the environment.

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b. Capability to recycle treated water in order to minimize the liquid radwaste effluent releases to the environment.

c. Capability to segregate the liquid waste streams by the use of separate waste drain headers and waste collection sumps or tanks for each waste stream category.

The LWMS provides sufficient capacity, redundancy, and flexibility to treat the liquid radwaste in a manner that reduces the radionuclide concentrations to levels that do not exceed the effluent concentration limits in 10 CFR Part 20, Appendix B, and 10 CFR Part 50, Appendix I (Reference 8) dose objectives for liquid effluents.

A description of this system is presented in Subsection 11.2.

1.2.13.2 Gaseous Waste Management System The gaseous waste management system (GWMS) is designed to monitor, control, collect, process, handle, store, and dispose of gaseous radioactive waste generated during normal plant conditions, including AOOs.

The GWMS manages radioactive gases collected from the off-gas system and other tank vents containing radioactive materials. The gaseous waste from the above sources is treated to reduce the quantity of radioactive material prior to release to the environment.

The radiation level in the processed gases is verified with radiation monitors prior to release to the environment.

The GWMS meets the following design requirements:

a. Provide the capability to monitor, control, collect, process, handle, store, and dispose of radioactive gaseous waste generated as the result of normal operation including AOOs to meet release radionuclide concentration limits in accordance with 10 CFR Part 20, Appendix B, prior to discharge to the environment.

b. Provide reasonable assurance that the release of radioactive material in gaseous effluents is kept as low as (is) reasonably achievable (ALARA).

c. Remove and reduce radioactive materials to the environment to meet the requirements of 10 CFR Part 50, Appendix I.

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APR1400 DCD TIER 2 The gaseous radwaste subsystem uses charcoal at ambient temperature to delay the passage of radioactive gases. When operating at design conditions, the mass of charcoal provided in the absorber beds is sufficient to provide a delay of 45 days for xenon and a delay of 3.5 days for krypton.

The GWMS operates at pressures slightly above atmospheric, and therefore limits the potential for oxygen inleakage. Leakage from the GWMS is further limited through the use of welded connections wherever they are not restricted due to maintenance requirements. Control valves are provided with bellows seals to minimize leakage through the valve stems. The GWMS is designed to prevent the formation or buildup of explosive mixtures of hydrogen and oxygen by continuous monitoring and controlling the concentrations of hydrogen and oxygen through one of the two gas analyzers. The concentrations are confirmed by periodic sampling and analysis at several routing locations.

When the oxygen concentration is detected to be higher than the predetermined setpoint, nitrogen is injected to dilute the concentration to below the lower flammable limit, which is 4 percent.

A description of this system is presented in Section 11.3.

1.2.13.3 Solid Waste Management System The solid waste management system (SWMS) is designed to provide the means to monitor, control, collect, process, handle, and temporarily store the following prior to shipment: wet, dewatered, and dry solid radioactive waste generated during normal plant conditions, including AOOs. The SWMS processes both wet solid active waste and dry active waste (DAW) for onsite interim storage and shipment to the offsite disposal facility.

The SWMS meets the following design requirements:

a. Collect, segregate, treat, package, and store various solid radioactive wastes generated from the normal operation, maintenance, refueling, and AOOs.

b. Store, treat, and package the radioactive spent resin transported from the LWMS, CVCS, SFPCCS, and steam generator blowdown system (SGBS).

c. Temporarily store the high- and low-activity waste, and to retrieve and ship wastes.

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d. Treat and package wastes into drums or high-integrity containers (HICs) that satisfy the required regulations of the U.S. Department of Transportation (DOT) and the disposal facility.

e. Satisfy federal regulations, and protect the workers and the general public from radiation exposures ALARA.

The SWMS is subdivided into a spent resin transfer subsystem, packaging and storage subsystem, filter handling subsystem, dry active waste subsystem, concentrate treatment subsystem, and waste storage subsystem.

In order to reduce occupational radiation exposure, operations for processing and transfer of low- and intermediate-level radioactive waste are conducted remotely. Operator access is required for work related to low-level radioactive waste such as DAW.

A description of this system is presented in Section 11.4.

1.2.14 Plant Arrangement Summary The APR1400 plant is composed of the following buildings:

a. Reactor containment building

b. Auxiliary building including two emergency diesel generator rooms

c. Turbine generator building

d. Compound building

e. Emergency diesel generator building with two emergency diesel generator rooms, including the DFOT room

f. Alternate alternating current gas turbine generator building

g. Essential service water intake structure and ultimate heat sink related structure

h. Component cooling water heat exchanger building

i. Security Buildings 1.2-35 Rev. 2

APR1400 DCD TIER 2 The scope of the certified design is shown on the typical site arrangement plan in Figure 1.2-1; the figure includes site specific structures and features. The general arrangement drawings are shown in Figures 1.2-2 through 1.2-49. For seismic analysis and structural design, the certified design scope is limited to the reactor containment building, the auxiliary building, and the emergency diesel generator building including the diesel fuel oil tank room, which are seismic Category I buildings. The other seismic Category I buildings, including the component cooling water heat exchanger building and the essential service water building, and the seismic Category II buildings, such as the turbine generator building, the compound building, and the AAC gas turbine generator building, are designed by the COL applicant.

1.2.14.1 Reactor Containment Building The reactor containment building is designed using a post-tensioned concrete containment wall with a reinforced concrete internal structure. The reactor containment building houses a reactor, two steam generators, a pressurizer, reactor coolant loops, an IRWST, and portions of the auxiliary systems. The reactor containment building is designed to provide biological shielding and external missile protection, as well as to sustain all internal and external loading conditions that are reasonably expected to occur during the life of the plant.

The interior arrangement of the reactor containment building is designed to meet the requirements for all anticipated conditions during operation and maintenance, including new and spent fuel handling.

The equipment hatch is located at the operating floor level. The hatch is sized to accommodate the one-piece replacement of a steam generator. A polar bridge crane is supported from the wall of the reactor containment building. The polar bridge crane has the capability to install and remove the steam generators. Personnel access from the auxiliary building to the reactor containment building is through two hatches: one at the operating floor and the other one at the ground floor.

1.2.14.2 Auxiliary Building The auxiliary building encompasses the reactor containment building and is on the common basemat that forms a monolithic structure with the reactor containment building. The auxiliary building houses the MCR, two EDG rooms, emergency core cooling system (ECCS) equipment area, fuel handling area, safety-related electrical and instrumentation and control (I&C) equipment areas, and two auxiliary feedwater storage tanks.

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APR1400 DCD TIER 2 The auxiliary building is designed as a seismic Category I reinforced concrete structure. It houses safety-related equipment required to provide safe shutdown capability. Redundant divisions of systems essential for safe shutdown are physically separated from one another to prevent a common failure of both systems.

The remote shutdown console (RSC) is located in a separate fire area from the MCR and contains all controls necessary for safe shutdown.

For the convenience of operation and maintenance, including for installation work, there is a staging service area in the auxiliary building in front of the equipment hatch of the reactor containment building.

The auxiliary building is physically separated into Division I and Division II. The divisions are subdivided with quadrant walls (Quadrant A through Quadrant D).

Quadrants A and C belong to Division I, and Quadrants B and D belong to Division II.

The APR1400 safety-related systems, including their components, are divided into the two divisions in the auxiliary building, which are physically and electrically independent of each other. The components are further divided into the four quadrants of the auxiliary building. A fire or flood in one quadrant does not affect the other quadrants.

The fuel handling area houses the following facilities:

a. Vehicle loading and unloading area A vehicle loading and unloading area is provided adjacent to the decontamination area. Space is provided within the building to permit inspection, tiedown adjustments, radiation monitoring, and storage of removable tiedown equipment.

Doors are sized to allow the traffic of personnel and the vehicles ingressing and exiting the building. A vehicle cleanup and maintenance area is provided outside the building to service incoming vehicles that includes removing road dirt, cinders, salt, oil, and similar materials.

b. Decontamination pit A decontamination pit is sized to permit the storage of a shipping cask, shipping cask head, and all other necessary rigging. Space also is allowed for portable scaffolds, elevated platforms, or ladders to gain access to the upper parts of the 1.2-37 Rev. 2

APR1400 DCD TIER 2 cask. Ample room is provided in the decontamination pit for the free passage of operating personnel around this equipment. Fixed or movable splash curtains or barriers are included to prevent splashing or accidental spillage out of the decontamination pit.

c. Spent fuel pool The spent fuel pool is designed to allow the installation of the number of spent fuel storage racks that are required to accommodate 20 years of discharged fuel plus one full core storage. The pool is designed to allow the installation of underwater lighting around the periphery of the pool to enhance visibility during fuel handling. Fuel assemblies are to be placed in vertical cells (storage racks) and grouped in parallel rows. Cranes used in moving spent fuel have a lift height limit in order to maintain the required water shielding above the spent fuel storage racks during transfer operations.

d. Overhead crane Spent fuel is loaded into a shipping cask in the cask loading pit using the spent fuel handling machine. The shipping cask is moved from the cask loading pit to the decontamination pit using an overhead crane. The cask travel path and the lifting height are restricted by limiting the distance above or adjacent to stored fuel within the spent fuel pool and the height of cask lift.

1.2.14.3 Turbine Generator Building The turbine generator building is designed as a seismic Category II steel frame building and has the following four main levels: a basement, a grade floor, an operating floor, and a deaerator and storage tanks floor. The basement consists of concrete and steel structures, and the above-grade floor consists of steel structures.

The turbine generator building provides support and housing for the turbine generator and auxiliary equipment. The generator-associated equipment includes condensers, feedwater heaters, feedwater and condensate pumps, and the condensate cleanup system. Auxiliary equipment includes the lube oil system, hydrogen supply and cooling system, stator cooling system, seal oil system, and electro-hydraulic control system.

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APR1400 DCD TIER 2 1.2.14.4 Compound Building The compound building houses the systems and components related to radwaste management, access control, and the operation support center (OSC). The compound building consists of an access control facility, radwaste management facility, hot machine shop, and sampling facilities and laboratory.

The compound building is adjacent to the auxiliary building, and is classified as non-safety-related seismic Category II reinforced concrete structures. The compound building is supported by a reinforced concrete foundation that is separated from the foundation of the auxiliary building and designed so that it will not affect safety-related structures system and components in the auxiliary building under the safe shutdown earthquake (SSE) condition.

The compound building is designed to be protected from natural phenomena such as flooding, snow, and earthquakes and to accommodate loadings associated with environmental conditions to the extent necessary to retain within the building the spillage of potentially contaminated solids or liquids.

1.2.14.5 Emergency Diesel Generator Building There are four EDG units. Two are located in the auxiliary building and two in the emergency diesel generator buildings (EDGBs). The two EDG units in the auxiliary building are separated so they are on opposite sides of the building in a mirror configuration.

The EDGB houses two EDG units and supporting equipment in two separate compartments.

The EDGB is a seismic Category I reinforced concrete structure and each EDG compartment is designed to be physically separate to provide protection from fire, aircraft, missiles, and the environment. The EDGB is also designed to withstand the effects of internal and external hazards.

1.2.14.6 Alternate Alternating Current Gas Turbine Generator Building The AAC gas turbine generator building is located on the north side of the plant site. The AAC gas turbine system provides an AAC power source during an SBO.

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APR1400 DCD TIER 2 1.2.14.7 Essential Service Water Building Two ESW buildings separated by division are classified as seismic Category I buildings with a concrete structure. The ESW building houses essential service water pumps, cooling tower, and cooling tower basin.

1.2.14.8 Component Cooling Water Heat Exchanger Building Two CCW heat exchanger buildings next to each ESW building are classified as a seismic Category I building with a concrete structure. The CCW heat exchanger building houses CCW heat exchangers, debris filters.

1.2.14.9 Security Buildings Two security buildings are located on the site to house security equipment and personnel.

Security Building 1 is located on the west side near transformer area and Security Building 2 is located near the Compound Building.

1.2.14.10 Storage Tanks The following storage tanks are located on the site, outside of building structures:

a. Reactor makeup water tank (seismic Category III)

b. Holdup tank (seismic Category III)

c. Boric acid storage tank (seismic Category I)

d. Condensate storage tank (seismic Category III)

e. Demineralized water storage tank (seismic Category III)

f. Fresh water storage tank (seismic Category III)

The reactor makeup water tank capacity is based on providing dilution to allow total recycle. The tank also provides dilution for one cold shutdown operation and subsequent startup at the most limiting time in core cycle. The reactor makeup water tank is described further in Subsection 9.3.4.

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APR1400 DCD TIER 2 The holdup tank is sized to store all recoverable reactor coolant generated by one cold shutdown operation with the most reactive CEA withdrawn and subsequent startup at the most limiting time in core cycle. The holdup tank is described further in Subsection 9.3.4.

The boric acid storage tank is sized to permit one shutdown operation to cold shutdown, followed by a shutdown for refueling at the most limiting time in core cycle with the most reactive CEA withdrawn. The maximum concentration of boric acid in the tank is 2.5 weight percent (4,400 ppm boron). The boric acid storage tank is described further in Subsection 9.3.4.

The fresh water storage tank is sized to contain sufficient water for 2-hour operation of the largest design demand of any sprinkler system plus a 1,900 L/min (500 gpm) manual hose stream allowance to support fire suppression activities, or at least 1,135,500 liters (300,000 gallons) in ground-level storage tanks. Two 100 percent fresh water storage tanks are arranged separately so that the fire pumps can take suction from either or both tanks. The fresh water storage tank is described further in Subsection 9.5.1.

One 100 percent capacity demineralized water storage tank is provided for APR1400. The demineralized water storage tank stores and supplies demineralized water to the auxiliary feedwater storage tank for makeup and to other systems for various services during all modes. The demineralized water storage tank is described further in Subsection 9.2.6.

Two 50 percent capacity condensate storage tanks store and supply the condensate, as a readily available source of deaerated condensate for makeup, to the condenser. The condensate storage tank is described further in Subsection 9.2.6.

1.2.15 Combined License Information COL 1.2(1) The COL applicant is to prepare a complete and detailed site plan.

1.2.16 References

1. 10 CFR 50.34, Contents of Applications; Technical Information, U.S. Nuclear Regulatory Commission.

2. 10 CFR Part 50, Appendix A, General Design Criterion 54, Systems Penetrating Containment, U.S. Nuclear Regulatory Commission.

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3. 10 CFR 50.62, Requirements for the Reduction of Risk from Anticipated Transients without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants, U.S.

Nuclear Regulatory Commission.

4. Staff Requirements Memorandum on SECY-93-087, Item II.Q, Defense Against Common-Mode Failures in Digital Instrumentation and Control Systems, U.S.

Nuclear Regulatory Commission, 1993

5. Regulatory Guide 1.97, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants, Rev. 4, U.S. Nuclear Regulatory Commission, June 2006.

6. 10 CFR Part 50, Appendix A, General Design Criterion 60, Control of Release of Radioactive Materials to the Environment, U.S. Nuclear Regulatory Commission.

7. 10 CFR Part 20, Appendix B, Annual Limits on Intake (ALIs) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage, U.S. Nuclear Regulatory Commission.

8. 10 CFR Part 50, Appendix I, Numerical Guides for Design Objectives and Limiting Conditions for Operation to meet the Criterion As Low As is Reasonably Achievable for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents, U.S. Nuclear Regulatory Commission.

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APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-1 Typical APR1400 Site Arrangement Plan 1.2-43 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-2 General Arrangement Reactor Containment Building Section A-A 1.2-44 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-3 General Arrangement Reactor Containment Building Section B-B 1.2-45 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-4 General Arrangement Reactor Containment Building El. 69-0 and El. 78-0 1.2-46 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-5 General Arrangement Reactor Containment Building El. 100-0 1.2-47 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-6 General Arrangement Reactor Containment Building El. 114-0 1.2-48 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-7 General Arrangement Reactor Containment Building El. 136-6 1.2-49 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-8 General Arrangement Reactor Containment Building El. 156-0 1.2-50 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-9 General Arrangement Auxiliary Building Section A-A 1.2-51 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-10 General Arrangement Auxiliary Building Section B-B 1.2-52 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-11 General Arrangement Auxiliary Building El. 55-0 1.2-53 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-12 General Arrangement Auxiliary Building El. 68-0 and El. 86-0 1.2-54 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-13 General Arrangement Auxiliary Building El. 78-0 1.2-55 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-14 General Arrangement Auxiliary Building El. 100-0 1.2-56 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-15 General Arrangement Auxiliary Building El. 120-0 1.2-57 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-16 General Arrangement Auxiliary Building El. 137-6 1.2-58 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-17 General Arrangement Auxiliary Building El. 156-0 1.2-59 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-18 General Arrangement Auxiliary Building El. 174-0 1.2-60 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-19 General Arrangement Auxiliary Building Roof El. 195-0 1.2-61 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-20 General Arrangement EDG Building El. 63-0, Sections A-A and B-B 1.2-62 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-21 General Arrangement EDG Building El. 100-0 and El. 121-6 1.2-63 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-22 General Arrangement EDG Building El. 135-0 1.2-64 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-23 General Arrangement Compound Building Sections A-A and B-B 1.2-65 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-24 General Arrangement Compound Building El. 63-0 1.2-66 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-25 General Arrangement Compound Building El. 77-0 1.2-67 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-26 General Arrangement Compound Building El. 85-0 1.2-68 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-27 General Arrangement Compound Building El. 100-0 1.2-69 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-28 General Arrangement Compound Building El. 120-0 1.2-70 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-29 General Arrangement Compound Building El. 139-6 1.2-71 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-30 General Arrangement Compound Building Roof El. 156-0 1.2-72 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-31 General Arrangement Turbine Generator Building Section A-A 1.2-73 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-32 General Arrangement Turbine Generator Building Section B-B 1.2-74 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-33 General Arrangement Turbine Generator Building El. 73-0 1.2-75 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-34 General Arrangement Turbine Generator Building El. 100-0 1.2-76 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-35 General Arrangement Turbine Generator Building El. 136-6 1.2-77 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-36 General Arrangement Turbine Generator Building El. 170-0 1.2-78 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-37 General Arrangement Turbine Generator Building Roof Plan 1.2-79 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-38 AAC Gas Turbine Generator Building Section A-A 1.2-80 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-39 AAC Gas Turbine Generator Building Plan El. 79-0 and El. 85-6 1.2-81 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-40 AAC Gas Turbine Generator Building Plan El. 100-0 1.2-82 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-41 AAC Gas Turbine Generator Building Plan El. 120-0 1.2-83 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-42 General Arrangement ESW/CCW Hx Building EL. 81-0 (DIV. I) 1.2-84 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-43 General Arrangement ESW/CCW Hx Building EL. 100-0 (DIV. I) 1.2-85 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-44 General Arrangement ESW/CCW Hx Building Roof Plan (DIV. I) 1.2-86 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-45 General Arrangement ESW/CCW Hx Building Section (DIV. I) 1.2-87 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-46 General Arrangement ESW/CCW Hx Building EL. 81-0 (DIV. II) 1.2-88 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-47 General Arrangement ESW/CCW Hx Building EL. 10-0 (DIV.II) 1.2-89 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-48 General Arrangement ESW/CCW Hx Building Roof Plan (DIV. II) 1.2-90 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 1.2-49 General Arrangement ESW/CCW Hx Building Section (DIV.II) 1.2-91 Rev. 2

APR1400 DCD TIER 2 1.3 Comparison with Other Facilities 1.3.1 Comparison with Similar Facility Designs This section highlights the principal features of the APR1400 design and provides a comparison of the major plant design features with other pressurized water reactor (PWR) facilities. Table 1.3-1 summarizes the comparison of design and operating characteristics for the nuclear steam supply system (NSSS) and demonstrates that the APR1400 NSSS design is similar to both a previously certified design and a design now nearing completion of construction in Korea. Table 1.3-2 summarizes major plant features other than those in the NSSS.

1.3-1 Rev. 2

APR1400 DCD TIER 2 Table 1.3-1 (1 of 13)

Comparison of NSSS Components DCD Tier 2 Design Feature APR1400 System 80+ SKN 3&4 Section Control Characteristics Dissolved boron content for criticality, ppm (CEAs withdrawn, BOC) 4.3 Cold, 20 °C (68 °F) 1,238 1,431 1,238 Hot, zero power, clean, 291.3 °C 1,187 1,414 1,187 (556.3 °F)

Hot, full power, equilibrium Xe 817 100 817 Hot, full power, clean, 308.9 °C 1,067 1,270 1,067 (588 °F)

Nuclear Design Data Structural Characteristics Core equivalent diameter, m (in) 3.647 (143.6) 3.647 (143.6) 3.647 (143.6) 4.1 Number of fuel assemblies 241 241 241 Core average H2O/UO2 volume ratio, 2.12 2.06 2.12 4.3 first cycle, hot (core cell)

UO2 fuel rod locations per assembly 236 (Batch A) 236 (1) 236 (Batch A) 236/224/220 - 236/224/220 (Batch B) (Batch B) 236/224/220 - 236/224/220 (Batch C) (Batch C) 1.3-2 Rev. 2

APR1400 DCD TIER 2 Table 1.3-1 (2 of 13)

DCD Tier 2 Design Feature APR1400 System 80+ SKN 3&4 Section Nuclear Design Data (cont.)

Performance Characteristics Fuel management 3 Batch Mixed 3 Batch Mixed 3 Batch Mixed 4.3 Central Zone Central Zone Central Zone Fuel discharge burnup, MWd/MtU Core average burnup, 10 ppm soluble 17,571 15,300 17,571 boron First core average discharge burnup 28,914 31,700 28,914 Fuel rod enrichment without U-235 Region 1 1.71 1.8 1.71 Region 2 3.14/2.64 2.9 3.14/2.64 Region 3 3.64/3.14 3.7 3.64/3.14 Region 4 - - -

Control Element Assemblies Material (full strength/part strength) B4C/Inconel B4C or Ag-In- B4C/Inconel 4.2 Cd/Inconel (2)

Number of control element assemblies 81/12 68/25 (3) 81/12 (full strength/part strength)

Number of absorber rods per CEA 4 or 12 4 or 12 4 or 12 (or rod cluster control assembly [RCCA])

Total rod worth (all CEAs inserted, 16.70 16.4 (typical) 16.70 4.3 hot, 308.9 °C (588 °F)), %

1.3-3 Rev. 2

APR1400 DCD TIER 2 Table 1.3-1 (3 of 13)

DCD Tier 2 Design Feature APR1400 System 80+ SKN 3&4 Section Nuclear Design Data (cont.)

Kinetic Characteristics Range Over First Cycle Moderator temperature coefficient -1.71x 10-4 / -1.3 x 10-4 / -1.71x 10-4 / 4.3

/°C (hot, full power, BOC/EOC) -4.34 x 10-4 -4.7 x 10-4 -4.34 x 10-4 Moderator pressure coefficient +0.44 x 10-6 +0.4 x 10-5 +0.44 x 10-6

/psi (hot, operating, BOC)

Moderator void coefficient /% -0.21 x 10-3 -0.22 x 10-3 -0.21 x 10-3 void (hot, operating, BOC)

Doppler coefficient /°C -2.54 x 10-5 / -2.74 x 10-5 / -2.54 x 10-5 /

(hot operating range, BOC/EOC) -2.95 x 10-5 -2.93 x 10-5 -2.95 x 10-5 Thermal and Hydraulic Design Parameters Total core heat output, MWt 3,983 3,914 3,983 4.4 6

Total core heat output, 10 kcal/hr 3,425 (13,590) 3,367 (13,360) 3,425 (13,590)

(MBtu/hr)

Average fuel rod energy deposition 0.975 0.975 0.975 fraction Primary system pressure, kg/cm2A (psia) 158.2 (2,250) 158.2 (2,250) 158.2 (2,250) 1.3-4 Rev. 2

APR1400 DCD TIER 2 Table 1.3-1 (4 of 13)

DCD Tier 2 Design Feature APR1400 System 80+ SKN 3&4 Section Engineering Factors Engineering heat flux factor 1.03 1.03 1.03 4.4 Engineering enthalpy rise factor 1.03 1.03 1.03 Engineering factor on LHR 1.03 1.03 1.03 Coolant Flow Total coolant flow, 106 kg/hr (106 lb/hr) 75.6 (166.6) 75.2 (165.8) 75.6 (166.6) 4.4 Core flow, 106 kg/hr (106 lb/hr) 73.3 (161.6) 73.0 (160.8) 73.3 (161.6)

Core flow area, m2 (ft2) 5.83 (62.7) 5.65 (60.8) 5.83 (62.7)

Core average coolant velocity, m/s (ft/s) 4.94 (16.2) 5.10 (16.7) 4.94 (16.2) 6 2 Core average mass velocity, 10 kg/hr-m 12.60 (2.58) 12.94 (2.65) 12.60 (2.58)

(106 lbm/hr-ft2) 1.3-5 Rev. 2

APR1400 DCD TIER 2 Table 1.3-1 (5 of 13)

DCD Tier 2 Design Feature APR1400 System 80+ SKN 3&4 Section Reactor Coolant Temperatures Reactor inlet coolant temperature, °C (°F) 290.6 (555) 291.1 (556) 290.6 (555) 4.4 (7)

Reactor outlet coolant temperature, °C 323.9 (615) 323.9 (615) 323.9 (615)

(°F) (7)

Core-exit average coolant temperature, °C 325.0 (617) 325.0 (617) 325.0 (617)

(°F) (7)

Average rise in vessel, °C (°F) (7) 33.3 (60) 32.8 (59) 33.3 (60)

Average rise in core, °C (°F) (7) 34.4 (62) 33.9 (61) 34.4 (62)

Average temperature in core, °C (°F) (7) 307.8 (586) 308.3 (587) 307.8 (586)

Average temperature in vessel, °C (°F) (7) 307.2 (585) 307.8 (586) 307.2 (585)

Characteristics of Rod and Channel with Minimum DNBR Outlet temperature, °C (°F) (7) 340.6 (645) 340.0 (644) 340.6 (645)

(4) (4)

Minimum DNBR at nominal conditions 2.44(KCE-1) 2.00(CE-1) 2.44(KCE-1)(4)

(CHF correlation) 1.3-6 Rev. 2

APR1400 DCD TIER 2 Table 1.3-1 (6 of 13)

DCD Tier 2 Design Feature APR1400 System 80+ SKN 3&4 Section Core Characteristics at Full Power Total heat transfer area, m2 (ft2) 6,454 (69,470) 6,592 (70,960) 6,454 (69,470) 4.4 Core average fuel rod heat flux, 517,361 497,200 517,361 kcal/hr-m2 (Btu/hr-ft2) (190,735) (183,300) (190,735)

Maximum fuel rod heat flux, 1,215,000 1,164,000 1,215,000 kcal/hr-m2 (Btu/hr-ft2) (448,000) (429,100) (448,000)

Average fuel rod LHR, W/cm (kW/ft) 179.2 (5.46) 175.9 (5.36) 179.2 (5.46)

Maximum fuel rod LHR, W/cm (kW/ft) 420.8 (12.8) 413.4 (12.6) 420.8 (12.8)

Maximum fuel centerline temperature at 1,712 (3,114) 1,748 (3,179) 1,712 (3,114) 100 % power, °C (°F)(7)

Mechanical Design Parameters Fuel Assemblies Fuel rod array square, 16 x 16 square, 16 x 16 square, 16 x 16 4.2 Fuel rod pitch, cm (in) 1.2852 (0.506) 1.2852 (0.506) 1.2852 (0.506)

Fuel rod to fuel rod, cm (in) 20.23 x 20.23 20.25 x 20.25 20.23 x 20.23 (7.964 x 7.964) (7.972 x 7.972) (7.964 x 7.964)

Total fuel weight, kg UO2 (lb UO2) 117.8 x 103 120.0x103 117.8 x 103 (assuming all rod locations are fuel (259.7 x 103) (264.5x103) (259.7 x 103) rods)

Number of grids per assembly 12 11 12 1.3-7 Rev. 2

APR1400 DCD TIER 2 Table 1.3-1 (7 of 13)

DCD Tier 2 Design Feature APR1400 System 80+ SKN 3&4 Section Fuel Rods Number of locations 56,876 (5) 56,876 (5) 56,876 (5) 4.1 Clad outside diameter (OD), cm (in) 0.950 (0.374) 0.970 (0.382) 0.950 (0.374) 4.2 Diametral gap (cold), cm (in) 0.01651 (0.0065) 0.01651 (0.0065) 0.01651 (0.0065)

Clad thickness, cm (in) 0.05715 (0.0225) 0.06350 (0.025) 0.05715 (0.0225)

Cladding material ZIRLO Zircaloy-4 ZIRLO Fuel Pellets Material UO2 Sintered UO2 Sintered UO2 Sintered 4.2 Diameter, cm (in) 0.8192 (0.3225) 0.827 (0.3255) 0.8192 (0.3225)

Length (enriched uranium), cm (in) 0.98 (0.387) 0.991 (0.390) 0.98 (0.387)

Control Element Assemblies Clad material Inconel 625 Inconel 625 Inconel 625 4.2 Clad thickness, cm (in) 0.089 (0.035) 0.089 (0.035) 0.089 (0.035) 1.3-8 Rev. 2

APR1400 DCD TIER 2 Table 1.3-1 (8 of 13)

DCD Tier 2 Design Feature APR1400 System 80+ SKN 3&4 Section Reactor Coolant System Code Requirements Reactor vessel ASME III, Class 1 ASME III, Class 1 ASME III, Class 1 5.2, 5.3, 5.4 Steam generator, tube side ASME III, Class 1 ASME III, Class 1 ASME III, Class 1 Steam generator, shellside ASME III, Class 2 ASME III, Class 2 ASME III, Class 2 Pressurizer ASME III, Class 1 ASME III, Class 1 ASME III, Class 1 Pilot operated safety relief valves ASME III, Class 1 ASME III, Class 1 ASME III, Class 1 Reactor coolant piping ASME III, Class 1 ASME III, Class 1 ASME III, Class 1 Principal Design Parameters of the Reactor Vessel Material Low-alloy, Low-alloy, Low-alloy, 5.3 steel with austenitic steel with austenitic steel with austenitic SS cladding SS cladding SS cladding Design pressure, kg/cm2A (psia) 175.8 (2,500) 175.8 (2,500) 175.8 (2,500)

Design temperature, °C (°F) 343.3 (650) 343.3 (650) 343.3 (650)

Normal operating pressure, kg/cm2A 158.2 (2,250) 158.2 (2,250) 158.2 (2,250)

(psia)

Inside diameter at shell, m (in) 4.63 (182-1/4) 4.63 (182-1/4) 4.63 (182-1/4)

Outside diameter across inlet nozzles, m 6.88 (271) 6.88 (271) 6.88 (271)

(in)

Overall height of vessel and head, m (in) to 14.83 (583-7/8) 14.83 (583-7/8) 14.83 (583-7/8) top (including closure head, CEDM nozzles and bottom head instrumentation nozzles)

Minimum cladding thickness, m (in) 0.318 (1/8) 0.318 (1/8) 0.318 (1/8) -

1.3-9 Rev. 2

APR1400 DCD TIER 2 Table 1.3-1 (9 of 13)

DCD Tier 2 Design Feature APR1400 System 80+ SKN 3&4 Section Principal Design Parameters of the Reactor Coolant Piping Material Carbon steel Carbon steel Carbon steel 5.4.3 internally clad with internally clad with internally clad with stainless steel stainless steel stainless steel Hot leg - ID, mm (in) 1,066.8 (42) 1,066.8 (42) 1,066.8 (42)

Cold leg - ID, mm (in) 762 (30) 762 (30) 762 (30)

Between pump and steam generator - ID, 762 (30) 762 (30) 762 (30) mm (in)

Design pressure, kg/cm2A (psia) 175.8 (2,500) 175.8 (2,500) 175.8 (2,500)

Principal Design Parameters of the Reactor Coolant System Operating pressure, kg/cm2A (psia) 158.2 (2,250) 158.2 (2,250) 158.2 (2,250) 5.1, 5.4 Reactor inlet temperature, °C (°F) (7) 290.6 (555) 291.1 (556) 290.6 (555)

Reactor outlet temperature, °C (°F) (7) 323.9 (615) 323.9 (615) 323.9 (615)

Number of loops 2 2 2 2

Design pressure, kg/cm A (psia) 175.8 (2,500) 175.8 (2,500) 175.8 (2,500)

Design temperature, °C (°F) 343.3 (650) 343.3 (650) 343.3 (650)

Hydrostatic test pressure (cold), 219.7 (3,125) 219.7 (3,125) 219.7 (3,125) kg/cm2A (psia)

Total coolant volume, m3 (ft3) (8) 453.4 (16,011) 448.1 (15,825.5) 453.4 (16,012)

Total reactor flow, L/min (gal/min) (6) 1,689,000 1,683,000 1,689,000 (446,300) (444,650) (446,300) 1.3-10 Rev. 2

APR1400 DCD TIER 2 Table 1.3-1 (10 of 13)

DCD Tier 2 Design Feature APR1400 System 80+ SKN 3&4 Section Nuclear Design Data Principal Design Parameters of the Reactor Coolant Pumps Number of units 4 4 4 -

Type Vertical, single-stage Vertical, single-stage Vertical, single-stage 5.4.1 centrifugal with centrifugal with centrifugal with bottom suction and bottom suction and bottom suction and horizontal discharge horizontal discharge horizontal discharge Design pressure, kg/cm2A (psia) 175.8 (2,500) 175.8 (2,500) 175.8 (2,500)

Design temperature, °C (°F) 343.3 (650) 343.3 (650) 343.3 (650)

Operating pressure, nominal, 158.2 (2,250) 158.2 (2,250) 158.2 (2,250) kg/cm2A (psia)

Suction temperature, °C (°F) (7) 290.6 (555) 291.1 (556) 290.6 (555)

Rated flow, L/min (gal/min) 460,256 (121,600) 437,014 (115,447) 460,256 (121,600)

Rated head, m (ft) 109.7 (360) 114.0 (374) 109.7 (360)

Hydrostatic test pressure, (cold) 219.7 (3,125) 219.7 (3,125) 219.7 (3,125) kg/cm2A (psia)

Motor type AC Induction AC Induction AC Induction Single Speed Single Speed Single Speed Motor rating (cold), kW (hp) 10,067 (13,500) 8,948 (12,000) 10,067 (13,500) 1.3-11 Rev. 2

APR1400 DCD TIER 2 Table 1.3-1 (11 of 13)

DCD Tier 2 Design Feature APR1400 System 80+ SKN 3&4 Section Nuclear Design Data Principal Design Parameters of the Steam Generators Number of units 2 2 2 5.4.2 Type Vertical U-tube with Vertical U-tube with Vertical U-tube with integral economizer integral economizer integral economizer Tube material SB-163 NiCrFe SB-163 NiCrFe SB-163 alloy 690 alloy 690 alloy 690 Shell material SA-533 Gr. B, Class Primary side low- SA-533 Gr. B, Class 1 or SA-508, Class 3 alloy steel clad with 1 or SA-508, Class 3 austenitic stainless steel Tube-side design pressure, kg/cm2A 175.76 (2,500) 175.76 (2,500) 175.76 (2,500)

(psia)

Tube-side design temperature, °C (°F) 343.33 (650) 343.33 (650) 343.33 (650) 6 6 Tube side design flow, kg/hr (lb/hr) per 37.78 x 10 (83.3 x 37.6 x 10 (82.9 x 37.78 x 106 (83.3 x steam generator 106) 106) 106)

Shellside design pressure, kg/cm2A 84.36 (1,200) 84.36 (1,200) 84.36 (1,200)

(psia)

Shellside design temperature, °C (°F) 298.88 (570) 298.88 (570) 298.88 (570)

Operating pressure, tube-side, nominal, 158.2 (2,250) 158.2 (2,250) 158.2 (2,250) kg/cm2A (psia)

Operating pressure, shellside, 77.3 (1,100) 77.3 (1,100) 77.3 (1,100) maximum, kg/cm2A (psia) 1.3-12 Rev. 2

APR1400 DCD TIER 2 Table 1.3-1 (12 of 13)

DCD Tier 2 Design Feature APR1400 System 80+ SKN 3&4 Section Maximum moisture at outlet at full 0.25 0.25 0.25 5.4.2 load, %

Hydrostatic test pressure, tube-side 219.7 (3,125) 219.7 (3,125) 219.7 (3,125) 5.2.2.1.2 (cold), kg/cm2A (psia)

Steam pressure, kg/cm2A (psia), at full 70.3 (1,000) 70.3 (1,000) 70.3 (1,000) 5.4.2 power Steam temperature, °C (°F) at full 285 (545) 285 (545) 285 (545) power Steam flow, at full power, lb/hr 4.070 x 106 4.0 x 106 4.070 x 106 per steam generator, kg/hr (lb/hr) (8.975 x 106) (8.82 x 106) (8.975 x 106) pressurizer Design pressure, kg/cm2A (psia) 175.8 (2,500) 175.8 (2,500) 175.8 (2,500) 5.4.10 Design temperature, °C (°F) 371.1 (700) 371.1 (700) 371.1 (700)

Normal operating pressure, kg/cm2A 158.2 (2,250) 158.2 (2,250) 158.2 (2,250)

(psia)

Normal operating temperature, °C (°F) 344.8 (652.7) 344.8 (652.7) 344.8 (652.7) 3 3 Internal free volume, m (ft ) 68.0 (2,400) 68.0 (2,400) 68.0 (2,400)

Normal (full power) operating water 33.2 (1,171) 34.0 (1,200) 33.2 (1,171) volume, m3 (ft3)

Normal (full power) steam volume, m3 35.7 (1,260) 34.9 (1,234) 35.7 (1,260)

(ft3)

Installed heater capacity, kW 2,400 2,400 2,400 1.3-13 Rev. 2

APR1400 DCD TIER 2 Table 1.3-1 (13 of 13)

DCD Tier 2 Design Feature APR1400 System 80+ SKN 3&4 Section ESFAS Initiating ESFAS Number of manual switches 2 sets of 2 each 2 sets of 2 each 2 sets of 2 each 7.3 Automatic initiation parameter 4 channels provided, 4 channels provided, 4 channels provided, channels/logic coincidence of 2 coincidence of 2 coincidence of 2 required for each required for each required for each function function function Initiating Reactor Trip Number of manual switches 2 sets of 2 in MCR 2 sets of 2 each in 2 sets of 2 in MCR 7.2 1 set of 2 in RSR both MCR and at 1 set of 2 in RSR RSP Automatic initiation parameter 4 channels provided, 4 channels provided, 4 channels provided, channels/ coincidence of coincidence of coincidence of logic 2 required for trip 2 required for trip 2 required for trip (1) In the first core, some UO2 rods may be replaced by burnable absorber rods.

(2) Inconel part-strength CEAs in System 80+

(3) Locations are provided for eight additional CEAs.

(4) Minimum DNBR at nominal conditions (5) Some of the rod locations are occupied by burnable absorber rods.

(6) Design minimum (7) Temperatures are given to the nearest degree.

(8) Cold condition including pressurizer 1.3-14 Rev. 2

APR1400 DCD TIER 2 Table 1.3-2 (1 of 5)

Comparison of Plant Components Other Than NSSS DCD Tier 2 Design Feature APR1400 System 80+ SKN 3&4 Section Containment System Containment Type Steel-lined Steel spherical Steel-lined 3.8, prestressed concrete containment shell, prestressed concrete 6.2 cylinder with surrounded by reinforced cylinder with hemispherical dome concrete shield building hemispherical dome Leak rate (%/d) 0.1 (24 hr) 0.5 (24 hr) 0.15 (24 hr) 0.05 (after 24 hr) 0.25 (after 24 hr) 0.075 (after 24 hr)

Design pressure, kg/cm2 (psig) 4.22 (60) 3.726 (53) 4.22 (60) 3 3 6 3 Net Free volume, 10 m (10 ft ) 88.576 (3.128) 95.626 (3.337) 88.576 (3.128)

Containment Spray 6.2.2 Number of pumps 2 2 2 Number of heat exchangers 2 2 2 Design capacity, each, lpm (gpm) 18,927 (5,000) 18,927 (5,000) 18,927 (5,000)

Containment Coolers 9.4.6 Type Normal and Normal and Normal and loss of offsite power loss of offsite power loss of offsite power Number of units 4 4 4 Capacity, Kcal/hr (Btu/hr) 1,725,264 755,988 (3,000,000) 1,725,264 (6,846,400) (6,846,400) 1.3-15 Rev. 2

APR1400 DCD TIER 2 Table 1.3-2 (2 of 5)

DCD Tier 2 Design Feature APR1400 System 80+ SKN 3&4 Section Electric System Onsite Power Systems, AC 8.3.1 Generator prime mover Diesel engine Diesel engine Diesel engine Number of units 4 2 2 Capacity, each (kW) 9,100 (EDG A, B) 5,500 8,000 7,500 (EDG C, D)

Other Systems Essential Service Water 9.2.1 Number of trains 2 2 2 Number of pumps/train 2 2 2 Pump type Vertical turbine Vertical centrifugal Vertical turbine type wet pit type wet pit type wet pit Rated flow rate, each, lpm (gpm) 75,708 (20,000) 54,889 (14,500) 64,352 (17,000)

Component Cooling Water 9.2.2 Number of trains 2 2 2 Number of pumps/train 2 2 2 Design capacity, each, lpm (gpm) 94,635 (25,000) 57,358 (15,200) 70,030 (18,500)

Number of heat exchangers/train 3 2 3 Heat exchanger type Plate Shell and tube Plate 1.3-16 Rev. 2

APR1400 DCD TIER 2 Table 1.3-2 (3 of 5)

DCD Tier 2 Design Feature APR1400 System 80+ SKN 3&4 Section Spent Fuel Pool Cooling and Cleanup 9.1.3 Cooling pump design capacity each, 15,142 (4,000) 13,249 (3,500) 15,142 (4,000) lpm (gpm)

Condensate Storage Facility 9.2.6 Total Capacity, L (103 gal) 1,930,560 (510,000) -- 1,930,560 (510,000)

Plant Fire Protection 9.5.1 Water source Fresh water tank Fresh water tank Fresh water tank Backup source Seismic Category I Seismic Category I Seismic Category I fire water storage fire water storage fire water storage tank tank tank Essential chilled water 9.2.7 Number of pump and chiller/divisions 2 (one per quad) 2 2 Emergency Diesel Generators 9.5.4 Fuel oil storage capacity per diesel 7 days, plus a margin 7 days, plus a margin 7 days, plus a margin operating at full power (days) for periodic testing for periodic testing for periodic testing Turbine Generator 10.2 Output, guaranteed (MWe) 1,425 1,391 1,455 1.3-17 Rev. 2

APR1400 DCD TIER 2 Table 1.3-2 (4 of 5)

DCD Tier 2 Design Feature APR1400 System 80+ SKN 3&4 Section Main Steam Supply 10.3, Total steam flow, kg/hr (1b/hr) 8.14x10 6 8.0x10 6 8.14x10 6 10.4.1 (17.95x106) (17.64x106) (17.95x106)

Steam generator dome pressure, 70.3 (1,000) 70.3 (1,000) 70.3 (1,000) kg/cm2 A (psia)

Steam generator dome temperature °C 284.2 (543.6) 284.8 (544.6) 284.8 (544.6)

(°F)

Condenser Type Single pressure Multi-pressure Single pressure Design Operating Pressure, kg/cm2A 0.09 (2.6) 0.06/0.08/0.10 0.05 (1.5)

(in HgA) (1.8/2.3/3.0)

Turbine Bypass 10.4.4 Capacity (% of rated load main steam 55 (to condenser) 55 (to condenser) 55 (to condenser) flow)

Auxiliary Feedwater 10.4.9 Pump prime movers 2 turbine-driven, 2 turbine-driven, 2 turbine-driven, 2 motor-driven 2 motor-driven 2 motor-driven Rated flow rate, each, L/min (gpm) 2,461 (650) 1,892 (500) 2,461 (650) 1.3-18 Rev. 2

APR1400 DCD TIER 2 Table 1.3-2 (5 of 5)

DCD Tier 2 Design Feature APR1400 System 80+ SKN 3&4 Section Radwaste Systems Liquid radwaste system tank volume, L 590,524 (156,000) (2) 2,175,491 (574,704) 635,949 (168,000) (2) 11.2 (1)

(gal) (one units) (two units)

(one unit)

Gaseous radwaste system holdup time 45 (xenon) 30 (xenon) 45 (xenon) 11.3 (days) 3.5 (krypton) 3 (krypton) 3.5 (krypton)

Solid radwaste system process type Solidification Dewatering Solidification 11.4 solidification agent polymer N/A polymer Emergency Support Facilities Technical support center (TSC) Dedicated TSC for Dedicated TSC for Dedicated TSC for 13.3 each unit, located in each unit, located in each unit, located in the auxiliary building the auxiliary building the auxiliary building of the respective unit of the respective unit of the respective unit (1) The neutralization tank is included (435,322 L (115,000 gal) x 2).

(2) APR1400 liquid radwaste systems have no neutralization tanks.

1.3-19 Rev. 2

APR1400 DCD TIER 2 1.4 Identification of Agents and Contractors The COL applicant that references the APR1400 design certification is to identify major agents, contractors, and participants for the construction and operation of the nuclear power plant (COL 1.4(1)).

1.4.1 Applicant - Program Manager Korea Electric Power Corporation (KEPCO) was founded with the objective to facilitate the development of electric power supply in Korea, meet the countrys power supply and demand needs, and contribute to the national economy in accordance with the Korea Electric Power Corporation Act. KEPCO is classified as a market-oriented public corporation under the Act on the Management of Public Institutions. KEPCOs areas of business are based on the aforementioned objectives and include the development of electric power resources, electric power generation, transmission, transformation, and distribution, as well as related marketing, research, technological development, overseas business, investment, corporate social responsibility, and use of its property. KEPCO provides funds for the APR1400 design certification and reviews top-level policy issues.

Korea Hydro & Nuclear Power (KHNP) is responsible for the APR1400 design. The design is based on the Korean standard nuclear reactor, the OPR1000. The OPR1000 design is based on more than 30 years of experience in the construction, operation, and design of nuclear power plants beginning with Kori Unit 1. Following the construction of Hanbit Units 3 & 4, which marked the beginning of the Korean standard nuclear power plant, KHNP has constructed and operated Hanbit Units 5 & 6, Hanul Units 3 through 6, and Shin-Kori Units 1 & 2. This application for design certification of the APR1400 design is also based on the ABB-CE System 80+ certified design. Using the design expertise accumulated while developing the OPR1000, KHNP upgraded the capacity for its standard design to 1,400 MW, making it a globally competitive reactor.

KHNP relies on the following three primary organizations for support of the APR1400 design certification: Korea Electric Power Corporation (KEPCO) Engineering &

Construction Company, Inc. (KEPCO E&C); KEPCO Nuclear Fuel Co., Ltd. (KEPCO NF);

and Doosan Heavy Industry & Construction Co., Ltd. (Doosan).

KEPCO is located in Seoul of the Republic of Korea, and KHNP is headquartered in Gyeongju. KEPCO E&C is headquartered in Yongin, KEPCO NF is located in Daejeon, and Doosan is located in Changwon.

1.4-1 Rev. 2

APR1400 DCD TIER 2 1.4.2 Architect Engineer (A/E) - KEPCO E&C KEPCO E&C is the prime contractor to KHNP for architectural and engineering services and other related services.

KEPCO E&C provides engineering and engineering management services, project management assistance, and support services.

KEPCO E&C was established in 1975 to meet the increasing demands for architectural and engineering capabilities in Korea. Since its inception, KEPCO E&C has played the leading role in consulting and engineering activities in all of the Korean nuclear projects.

The nuclear experience of KEPCO E&C dates to 1976, when it undertook several design tasks on the first nuclear power plant in Korea. Subsequently, KEPCO E&C has participated in the construction projects of all Korean nuclear power plants and has provided various engineering services. Recently, KEPCO E&C has been actively expanding business areas that include international projects.

1.4.3 Major Equipment Supplier - DOOSAN DOOSAN was established in 1962 under the name of Hyundai International Inc., and its primary business was the manufacture of various industrial machinery and equipment.

In the course of its development, the company changed management in November 1980 in conformance with the Korea Governmental Policy on Heavy Industry Distribution and was renamed Korea Heavy Industries and Construction Co., Ltd. (HANJUNG).

The structural reform of electric equipment was completed in November 1999 based on the policy as to the management structural improvement of public enterprise in April 1998.

HANJUNG was unificated in the division of electric equipment. Under the privatization policy of public enterprise, HANJUNG was renamed DOOSAN in March 2001 in its initial public offering in October 2000.

This government action and subsequent transformation in the ownership of the company has strengthened DOOSANs capabilities to assume sole responsibility for all nuclear power plants to be installed in Korea.

Under the above policy, DOOSAN was designated by KHNP as the prime contractor for the supply of equipment, materials, and related services of the nuclear steam supply system 1.4-2 Rev. 2

APR1400 DCD TIER 2 (NSSS) and turbine generators (T/Gs) for HUN 3&4, HBN 5&6, HUN 5&6, Shin-Kori 1&2, Shin-Wolsong 1&2, and Shin-Kori 3&4 in Korea.

Currently, DOOSAN is committed to the manufacturing of NSSSs and T/Gs for Shin-Hanul 1&2 in Korea.

1.4.4 Nuclear Steam Supply System Designer - KEPCO E&C KEPCO E&C - Nuclear Steam Supply System (KEPCO E&C-NSSS) division worked with Asea Brown-Boveri Combustion Engineering (ABB-CE) on the 1,000 MW pressurized water reactor (PWR) Hanbit Units 3 & 4 in 1987 and was able to develop the technology to achieve self-reliance of the NSSS design.

Since 1991, KEPCO E&C-NSSS has been designing the NSSS of all nuclear power plants built in Korea, including the 1,000 MW PWR, and has developed the NSSS design of the 1,400 MW APR1400 plant.

1.4.5 Nuclear Fuel Design and Manufacturing - KEPCO NF KEPCO NF is a fuel design and fabrication company that has been responsible for the fuel supplied to all nuclear power plants in Korea for decades. Its major activities include initial and reload core design, fuel development, fuel assembly and component manufacture, and fuel services.

1.4.6 Combined License Information COL 1.4(1) The combined license (COL) applicant that references the APR1400 design certification is to identify major agents, contractors, and participants for the construction and operation of the nuclear power plant.

1.4-3 Rev. 2

APR1400 DCD TIER 2 1.5 Requirements for Additional Technical Information This section describes additional technical information for the unique design features of the APR1400.

1.5.1 Fluidic Device Design Conventional nuclear power plants are designed to deliver cooling water into a reactor vessel from safety injection tanks (SITs) in the refill phase and to deliver the cooling water by safety injection pumps (SIPs) during the reflood phase in the event of a large-break loss-of-coolant accident (LBLOCA). During an LBLOCA, the fuel cladding temperature increases because the liquid around the core is carried away by a significant loss of reactor coolant from the RCS.

The safety injection system (SIS) is required to inject cooling water into the core to limit the fuel temperature increase. Safety injection (SI) water in the SIT plays the role of rapidly raising the water level in the downcomer up to the cold leg bottom in the refill stage, removing decay heat of the reactor core and sensible heat of the nuclear fuels and metal structures in the early reflood phase. The SITs of conventional nuclear power plants deliver excessive cooling water to the reactor vessel after the water level has been raised to the cold leg bottom elevation, causing SI water to flow into the containment atmosphere.

The excess flow to the containment limits the usefulness of SI water and can cause a decrease in the reflood rate.

The fluidic device (FD) is installed inside the pressurized SIT of the APR1400 and passively controls the water injection flow rate from the SIT. The FD consists primarily of a standpipe and vortex chamber. The vortex chamber can receive injection water through control ports as well as through the standpipe. When the SIT water level is above the top of the standpipe, water enters the vortex chamber through both the top of the standpipe and the control ports, resulting in injection of water at a large flow rate. When the water level drops below the top of the standpipe, water only enters the vortex chamber through the control ports, resulting in vortex formation in the vortex chamber and a relatively small flow injection. Therefore, the SIT provides short-term large flow injection to refill the reactor vessel, and a smaller flow injection, which, in conjunction with the SIP, adequately supports the core reflooding phase. The FD improves the LOCA safety analysis by reducing the amount of SI water that would spill into the containment once the downcomer is full, thereby improving the overall reliability of SI water injection.

1.5-1 Rev. 2

APR1400 DCD TIER 2 The performance of the FD has also been evaluated by repeated experiments in the full-scale valve performance evaluation rig (VAPER) Test Facility at the Korea Atomic Energy Research Institute (KAERI). The experimental results confirm that the currently developed FD satisfies the major performance requirements of the plant design regarding the injection flow rate, pressure loss coefficient (K-factor), and injection duration time.

The pressure loss coefficient of the small flow rate period is almost 10 times higher than that of the large flow rate period due to the strong vortex motion in the FD. The K-factor of the SIS has been evaluated based on the K-factor obtained from the tests, and this value essentially matches the target design value of the SIS in the APR1400.

A quality assurance program for the FD tests has been applied to provide reasonable assurance of high-quality documentation of the testing.

1.5.2 Pilot Operated Safety Relief Valve Design The APR1400 adapts the pilot operated safety relief valve (POSRV) to provide overpressure protection of the reactor coolant system (RCS). Four POSRVs are connected to the top of the pressurizer by separate inlet lines. These valves also provide rapid depressurization functions during the beyond-design-basis event of a total loss of feedwater event for feed-and-bleed operations and for severe accidents to reduce RCS pressure prior to vessel breach.

The four POSRVs are designed to maintain the RCS pressure below 110 percent of design pressure during the worst-case transient, a loss-of-load event with a delayed reactor trip.

The pressurizer POSRVs perform the overpressure protection function with two pairs of spring-loaded pilot assemblies and a main valve and perform the rapid depressurization function using two motor-operated pilot valves in series and a main valve.

Each pressurizer POSRV provides the overpressure protection function with a main valve and two spring-loaded pilot valves in assembly. Each spring-loaded pilot valve in the assembly consists of a motor-operated isolation valve, spring-loaded pilot valve, check valve, and manual isolation valve. The spring-loaded pilot valve of each POSRV opens automatically if the system pressure increases to the POSRV set pressure, thus opening the main valve. The motor-operated isolation valves are normally open but are manually closed by an operator to prevent discharge when the spring-loaded pilot valves fail to close.

The manual isolation valves are normally open but are manually closed to block the main valve from opening when repairing a spring-loaded pilot valve or conducting a setpoint test.

1.5-2 Rev. 2

APR1400 DCD TIER 2 Each pressurizer POSRV provides the rapid depressurization function with a main valve and two motor-operated pilot valves installed in series. The motor-operated pilot valves are normally closed, but an operator remotely and manually opens the valves to open the main valve for the rapid depressurization of the RCS.

The pressurizer POSRVs are capable of discharging steam, water, and steam-water mixture.

A typical POSRV is illustrated in Figure 5.4.14-1. The design parameters are given in Table 5.4.14-1.

1.5.3 Direct Vessel Injection One of the advanced design features in the safety injection system (SIS) for the APR1400 is direct vessel injection (DVI), which reduces the loss of injection water during cold leg pipe breaks. The SIS is designed to provide DVI. The discharge of each safety injection (SI) pump and safety injection tank (SIT) is piped directly to a reactor vessel nozzle instead of a cold leg nozzle. The flow is directed into the reactor vessel downcomer region through the DVI nozzle. The APR1400 design places the four DVI nozzles on the reactor vessel above the hot leg / cold leg centerline at angles of 45°, 135°, 225°, and 315° from the referenced hot leg.

1.5.4 Instrumentation and Control System The APR1400 instrumentation and control (I&C) systems consist of the safety I&C system, non-safety control and monitoring system, diverse actuation system (DAS), and human-system interface (HSI) system.

The safety I&C system consists of the plant protection system (PPS), core protection calculator system (CPCS), engineered safety features - component control system (ESF-CCS), and the qualified indication and alarm system - P (QIAS-P). The control and monitoring system includes the power control system (PCS), process-component control system, qualified indication and alarm system - non-safety (QIAS-N), and information processing system (IPS). The DAS is composed of the diverse protection system (DPS),

diverse indication system (DIS), and diverse manual ESF actuation (DMA) switch. The HSI system includes the compact workstation-based operator console with an information flat panel display and ESF-CCS soft control module (ESCM), large display panel, safety console with ESCM / manual switches / operator module / display device in the main 1.5-3 Rev. 2

APR1400 DCD TIER 2 control room, compact workstation-based operator console with ESCM, and a shutdown overview panel in the remote shutdown room.

The safety I&C system is implemented on the four channels of common programmable logic controller qualified for Class 1E grade in accordance with IEEE Std. 603 (Reference 1) and IEEE Std. 7-4.3.2 (Reference 2), and each channel is located in the separate I&C equipment room.

The software for the digital I&C system is designed, verified, and validated in accordance with software life-cycle process conforming with NRC RG 1.152 (Reference 3).

The control and monitoring system is implemented on a distributed control system.

The diversity and defense-in-depth analysis is performed to demonstrate that the DAS and control system meet the SRM on SECY-93-087, Item II.Q (Reference 4) in case of software common-cause failure in the safety I&C system. The DAS is implemented on the platform diverse from the safety I&C system and control system.

The data communication system provides a high-speed and error-free communication path between each system and within a system.

The HSI system is designed in accordance with the human factors engineering (HFE) program to provide reasonable assurance that the HFE design is properly developed and effectively implemented. The HFE program objectives for the NPP design are that the design is human-centered, it incorporates HFE principals and methods, and is developed according to a systematic top-down approach. In accordance with applicable requirements of the HFE process elements, the HFE program plan provides reasonable assurance that the HSI design effectively supports the operator and allows consequential operator errors to be minimized. The HFE program is in effect at least from the start of the design cycle through completion of initial plant startup test program to conform with NUREG-0711 (Reference 5).

1.5.5 References

1. IEEE Std. 603-1991, IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 1991.

1.5-4 Rev. 2

APR1400 DCD TIER 2

2. IEEE Std. 7-4.3.2-2003, IEEE Standard Criteria for Digital Computers in Safety Systems of Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 2003

3. Regulatory Guide 1.152, Criteria for Use of Computers in Safety Systems of Nuclear Power Plants, Rev. 3, U.S. Nuclear Regulatory Commission, July 2011.

4. SRM on SECY-93-087, Item II.Q, Defense against Common-Mode Failures in Digital Instrumentation and Control Systems, U.S. Nuclear Regulatory Commission, July 1993.

5. NUREG-0711, Human Factors Engineering Program Review Model, Rev. 3, U.S.

Nuclear Regulatory Commission, November 2012.

1.5-5 Rev. 2

APR1400 DCD TIER 2 1.6 Material Referenced Tables 1.6-1 and 1.6-2 contain lists of topical reports and technical reports, respectively, that are incorporated by reference in this document.

1.6-1 Rev. 2

APR1400 DCD TIER 2 Table 1.6-1 List of Topical Reports DCD Tier 2 (1)

Report Number Title Section APR1400-F-A-TR-12004-P Realistic Evaluation Methodology for 6.2.1.5.1, 15.6 APR1400-F-A-TR-12004-NP Large-Break LOCA of the APR1400, Rev. 1 APR1400-F-C-TR-12002-P-A KCE-1 Critical Heat Flux Correlation 4.4, APR1400-F-C-TR-12002-NP-A for PLUS7 Thermal Design, Rev. 0 15.0~15.6 APR1400-F-M-TR-13001-P PLUS7 Fuel Design for the APR1400, 4.2, 4.4 APR1400-F-M-TR-13001-NP Rev. 1 APR1400-K-Q-TR-11005-NP KHNP QAPD for the APR1400 DC, 17.1, 17.2, 17.3, Rev. 5 17.5 APR1400-Z-M-TR-12003-P-A Fluidic Device Design for the 6.3.2.1 APR1400-Z-M-TR-12003-NP-A APR1400, Rev. 0 WCAP-10697-P-A Common Qualified Platform Topical 7.1 Report, Rev. 3 (1) P - denotes document is proprietary.

NP - denotes document is non-proprietary.

1.6-2 Rev. 2

APR1400 DCD TIER 2 Table 1.6-2 (1 of 4)

List of Technical Reports DCD Tier 2 (1)

Report Number Title Section APR1400-E-A-NR-14002-P-SGI Physical Security Design Features 13.6.2 APR1400-E-I-NR-14001-P Human Factors Engineering Program 18.1 APR1400-E-I-NR-14001-NP Plan APR1400-E-I-NR-14002-P Operating Experience Review 18.2 APR1400-E-I-NR-14002-NP Implementation Plan APR1400-E-I-NR-14003-P Functional Requirements Analysis 18.3 APR1400-E-I-NR-14003-NP and Function Allocation Implementation Plan APR1400-E-I-NR-14004-P Task Analysis Implementation Plan 18.4 APR1400-E-I-NR-14004-NP APR1400-E-I-NR-14006-P Treatment of Important Human 18.6 APR1400-E-I-NR-14006-NP Actions Implementation Plan APR1400-E-I-NR-14007-P Human-System Interface Design 18.7 APR1400-E-I-NR-14007-NP Implementation Plan APR1400-E-I-NR-14008-P Human Factors Verification and 18.10 APR1400-E-I-NR-14008-NP Validation Implementation Plan APR1400-E-I-NR-14011-P Basic Human-System Interface 18.7 APR1400-E-I-NR-14011-NP APR1400-E-I-NR-14012-P Style Guide 18.7 APR1400-E-I-NR-14012-NP APR1400-E-N-NR-14001-P Design Features to Address GSI-191 6.2.1.1.2.2 APR1400-E-N-NR-14001-NP 6.8.2.2.1 APR1400-E-P-NR-14005-P Evaluations and Design 1.9.6, 19.3 APR1400-E-P-NR-14005-NP Enhancements to Incorporate Lessons Learned from FUKUSHIMA DAI-CHI Nuclear Accident APR1400-E-S-NR-14004-P Evaluation of Effects of HRHF 3.7B.1 APR1400-E-S-NR-14004-NP Response Spectra on SSCs 1.6-3 Rev. 2

APR1400 DCD TIER 2 Table 1.6-2 (2 of 4)

DCD Tier 2 Report Number (1) Title Section APR1400-E-S-NR-14005-P Evaluation of Structure-Soil-Structure 3.7.2.8 APR1400-E-S-NR-14005-NP Interaction (SSSI) Effects APR1400-E-S-NR-14006-P Stability Check for NI Common 3.8.5.4.3 APR1400-E-S-NR-14006-NP Basemat APR1400-F-A-NR-14001-P Small Break LOCA Evaluation Model 15.6 APR1400-F-A-NR-14001-NP APR1400-F-A-NR-14002-P The Effect of Thermal Conductivity 15.4 APR1400-F-A-NR-14002-NP Degradation on APR1400 Design and 15.6 Safety Analyses APR1400-F-A-NR-14003-P Post-LOCA Long Term Cooling 15.6 APR1400-F-A-NR-14003-NP Evaluation Model APR1400-H-N-NR-14012-P Mechanical Analysis for New and 9.1.2 APR1400-H-N-NR-14012-NP Spent Fuel Storage Racks APR1400-K-I-NR-14005-P Staffing and Qualifications 18.5 APR1400-K-I-NR-14005-NP Implementation Plan APR1400-K-I-NR-14009-P Design Implementation Plan 18.11 APR1400-K-I-NR-14009-NP APR1400-Z-A-NR-14006-P Non-LOCA Safety Analysis 15.0.2 APR1400-Z-A-NR-14006-NP Methodology APR1400-Z-A-NR-14007-P Mass and Energy Release 6.2.1.3 APR1400-Z-A-NR-14007-NP Methodologies for LOCA and MSLB APR1400-Z-A-NR-14011-P Criticality Analysis of New and Spent 9.1.1 APR1400-Z-A-NR-14011-NP Fuel Storage Racks APR1400-A-N-NR-17001-P Validation of SCALE 6.1.2 with 9.1.1 (WCAP-17889-P) 238-Group ENDF/B-VII.0 Cross APR1400-A-N-NR-17001-NP Section Library for APR1400 Design (WCAP-17889-NP) Certification APR1400-Z-J-NR-14001-P Safety I&C System 7.1, 7.2, 7.3, 7.4, APR1400-Z-J-NR-14001-NP 7.5, 7.8, 7.9 APR1400-Z-J-NR-14003-P Software Program Manual 7.1.4, 7.2.2.2, APR1400-Z-J-NR-14003-NP 7.3.1 APR1400-E-J-NR-17001-P Secure Development and Operational 7.1.2.46 APR1400-E-J-NR-17001-NP Environment for APR1400 Computer-Based I&C Safety Systems 1.6-4 Rev. 2

APR1400 DCD TIER 2 Table 1.6-2 (3 of 4)

DCD Tier 2 Report Number (1) Title Section APR1400-Z-J-NR-14004-P Uncertainty Methodology and 7.2.2.7, 7.3.2.7 APR1400-Z-J-NR-14004-NP Application for Instrumentation APR1400-Z-J-NR-14005-P Setpoint Methodology for Safety- 7.2.2.7, 7.3.2.7 APR1400-Z-J-NR-14005-NP Related Instrumentation APR1400-A-J-NR-14003-P APR1400 Disposition of Common Q 7.1 (WCAP-17926-P) Topical Report NRC Generic Open Items and Plant Specific Action Items APR1400-A-J-NR-14004-P Common Q Platform Supplemental 7.1 (WCAP-17922-P) Information in Support of the APR1400 Design Certification APR1400-E-J-NR-14001-P Component Interface Module 7.3.1.11, 7.3.2.4, APR1400-E-J-NR-14001-NP 7.8 APR1400-F-C-NR-14003-P Functional Design Requirements for a 7.2.1.1 APR1400-F-C-NR-14003-NP Core Protection Calculator System for APR1400 APR1400-Z-A-NR-14019-P CCF Coping Analysis 7.7.2.8, 7.8 APR1400-Z-A-NR-14019-NP APR1400-Z-J-NR-14002-P Diversity and Defense-in-Depth 7.1, 7.2, 7.3, 7.7, APR1400-Z-J-NR-14002-NP 7.8 APR1400-Z-J-NR-14012-P Control System CCF Analysis 7.7.2.3, 7.7.2.8 APR1400-Z-J-NR-14012-NP APR1400-Z-J-NR-14013-P Response Time Analysis of Safety 7.2.2.7, 7.3.2.7 APR1400-Z-J-NR-14013-NP I&C System APR1400-Z-M-NR-14008-P Pressure-Temperature Limits 5.2, 5.3 APR1400-Z-M-NR-14008-NP Methodology for RCS Heatup and Cooldown APR1400-F-C-NR-14001-P CPC Setpoint Analysis Methodology 7.2.2.7 APR1400-F-C-NR-14001-NP for APR1400 APR1400-F-C-NR-14002-P Functional Design Requirements for a 7.7.1.4 APR1400-F-C-NR-14002-NP Core Operating Limit Supervisory System for APR1400 1.6-5 Rev. 2

APR1400 DCD TIER 2 Table 1.6-2 (4 of 4)

DCD Tier 2 Report Number (1) Title Section APR1400-E-B-NR-16001-P Evaluation of Main Steam and 3.12 APR1400-E-B-NR-16001-NP Feedwater Piping applied to the Graded Approach for the APR1400 APR1400-E-B-NR-16002-P Evaluation of Safety Injection and 3.12 APR1400-E-B-NR-16002-NP Shutdown Cooling Piping Applied to the Graded Approach for the APR1400 APR1400-H-N-NR-14005-P Summary Stress Report for Primary 3.12 APR1400-H-N-NR-14005-NP Piping CEN-310-P-A CPC and Methodology Changes for 7.2 the CPC Improvement Program CEN-312-P Overview Description of the Core 7.7 Operating Limit Supervisory System APR1400-E-X-NR-14001-P Equipment Qualification Program 3.10, 3.11 APR1400-E-X-NR-14001-NP (1) P - denotes document is proprietary.

NP - denotes document is non-proprietary.

1.6-6 Rev. 2

APR1400 DCD TIER 2 1.7 Drawings and Diagrams 1.7.1 Electrical, Instrumentation, and Control Drawings The line drawings of offsite and onsite electrical systems are provided in Chapter 8.

Instrumentation, control, and other electrical drawings are provided in Chapter 7. Table 1.7-1 is a list of the electrical, instrumentation, and control drawings in Chapters 7 and 8.

1.7.2 Flow Diagrams The flow diagrams are listed in Table 1.7-2. The symbols used in the flow diagrams are defined in Figure 1.7-1.

1.7-1 Rev. 2

APR1400 DCD TIER 2 Table 1.7-1 (1 of 4)

Safety-Related Electrical, Instrumentation, and Control Drawings DCD Tier 2 Figure No. Title Section 7.1-1 APR 1400 I&C System Overview Architecture 7.1 7.1-2 Symbol & Legend Diagram 7.1 7.2-1 PPS Basic Block Diagram 7.2 7.2-2 PPS Measurement Channel Functional Diagram 7.2 (Pressurizer Pressure Narrow Range) 7.2-3 Reed Switch Position Transmitter Assembly Schematic 7.2 7.2-4 CEA Position Signal Flow for CPCS 7.2 7.2-5 Ex-Core Neutron Monitoring System (Safety Channel) 7.2 7.2-6 Reactor Coolant Pump Shaft Speed Sensing System 7.2 7.2-7 Core Protection Calculator System Functional Block Diagram 7.2 7.2-8 PPS Bistable Trip Logic Functional Block Diagram 7.2 7.2-9 Reactor Trip Switchgear System Interface Diagram 7.2 7.2-10 PPS Channel A Trip Path Diagram 7.2 7.2-11 PPS Testing Overlap 7.2 7.2-12 Interface and Test Processor Block Diagram 7.2 7.2-13 PPS Channel Contact Bistable Interface Diagram 7.2 7.2-14 Plant Protection System Interface Logic Diagram for Division D 7.2 7.2-15 Reactor Trip Initiation Diagram 7.2 7.2-16 Manual Reactor Trip Initiation Diagram 7.2 7.2-17 Functional Logic Diagram for Variable Overpower 7.2 7.2-18 Functional Logic Diagram for High Logarithmic Power Level 7.2 7.2-19 Functional Logic Diagram for High Local Power Density 7.2 7.2-20 Functional Logic Diagram for Low Departure from Nucleate 7.2 Boiling Ratio 7.2-21 Functional Logic Diagram for High Pressurizer Pressure 7.2 7.2-22 Functional Logic Diagram for Low Pressurizer Pressure 7.2 7.2-23 Functional Logic Diagram for Low Steam Generator Water Level 7.2 7.2-24 Functional Logic Diagram for Low Steam Generator Pressure 7.2 1.7-2 Rev. 2

APR1400 DCD TIER 2 Table 1.7-1 (2 of 4)

DCD Tier 2 Figure No. Title Section 7.2-25 Functional Logic Diagram for High Containment Pressure 7.2 7.2-26 Functional Logic Diagram for High Steam Generator Water 7.2 Level 7.2-27 Functional Logic Diagram for Low Reactor Coolant Flow 7.2 7.2-28 Functional Logic Diagram for Reactor Trip Signal Generation 7.2 7.2-29 Functional Logic Diagram for DNBR/LPD Operating Bypass 7.2 Permissive 7.2-30 Functional Logic Diagram for Low Pressurizer Pressure 7.2 Operating Bypass Permissive 7.2-31 Functional Logic Diagram for High Logarithmic Power Level 7.2 Operating Bypass Permissive 7.2-32 Functional Logic Diagram for CWP 7.2 7.3-1 Simplified Functional Diagram of the ESF-CCS 7.3 7.3-2 Block Diagram of the ESF-CCS 7.3 7.3-3 ESF-CCS Simplified Logic Diagram for 2-out-of-4 Actuation 7.3 7.3-4 ESFAS Functional Logic (SIAS) 7.3 7.3-5 ESFAS Functional Logic (CSAS) 7.3 7.3-6 ESFAS Functional Logic (CIAS) 7.3 7.3-7 ESFAS Functional Logic (MSIS) 7.3 7.3-8 ESFAS Functional Logic (AFAS-1 and AFAS-2) 7.3 7.3-9 ESFAS Functional Logic (FHEVAS) 7.3 7.3-10 ESFAS Functional Logic (CPIAS) 7.3 7.3-11 ESFAS Functional Logic (CREVAS) 7.3 7.3-12 CLD for a Solenoid-Operated Valve 7.3 7.3-13 CLD for a Modulating Valve with Solenoid Operator 7.3 7.3-14 Motor-Operated Valve Functional Interface Design 7.3 7.3-15 CLD for a Full Stroke Motor-Operated Valve 7.3 7.3-16 CLD for a Throttling Motor-Operated Valve 7.3 7.3-17 CLD for a Non-reversing Motor Starter Operated Component 7.3 7.3-18 CLD for a Circuit Breaker Operated Component 7.3 7.3-19 CLD for a Modulating Component 7.3 1.7-3 Rev. 2

APR1400 DCD TIER 2 Table 1.7-1 (3 of 4)

DCD Tier 2 Figure No. Title Section 7.3-20 CLD for a Electro-Hydraulic Motor Damper 7.3 7.3-21 EDG Loading Sequencer - Control Logic Diagram 7.3 7.3-22 ESF-CCS Simplified Test Logic Diagram 7.3 7.3-23 Radiation Monitoring System Measurement Channel Functional 7.3 Diagram 7.4-1 Interface Diagram for Division A MCR/RSR Master Transfer 7.4 Switches 7.4-2 Interface Diagram for Division AB MCR/RSR Master Transfer 7.4 Switches 7.4-3 MCR/RSR Master Transfer Logic (Division A) 7.4 7.4-4 Layout of Remote Shutdown Room 7.4 7.5-1 Diverse Display of Accident Monitoring Type B and C Variables 7.5 7.5-2 QIAS-N Block Diagram 7.5 7.6-1A Interlocks for Shutdown Cooling System Suction Line Isolation 7.6 Valve 7.6-1B Interlocks for Shutdown Cooling System Suction Line Isolation 7.6 Valve 7.6-1C Interlocks for Shutdown Cooling System Suction Line Isolation 7.6 Valve 7.6-2 Interlocks for Safety Injection Tank Isolation Valve 7.6 7.6-3 Interlocks for CCW Supply and Return Header Isolation Valve 7.6 7.7-8 Process-Component Control System Simplified Block Diagram 7.7 7.7-11 N-16 Detection and Alarm Logic 7.7 7.7-12 HSI Information Processing Block Diagram 7.7 7.7-13 Layout of Main Control Room 7.7 7.8-1 Diverse Protection System Block Diagram 7.8 7.8-2 Diverse Reactor Trip, Turbine Trip, AFWS and SIS Actuation 7.8 7.8-3 Diverse Reactor Trip and Turbine Trip 7.8 7.8-4 Diverse AFWS Actuation 7.8 7.8-5 Diverse SIS Actuation 7.8 7.8-6 DMA Switches Block Diagram 7.8 1.7-4 Rev. 2

APR1400 DCD TIER 2 Table 1.7-1 (4 of 4)

DCD Tier 2 Figure No. Title Section 7.9-1 Data Communication Block Diagram 7.9 8.1-1 Electric Power System Single Line Diagram 8.1 8.2-1 Layout Drawing Showing MT, UAT, SAT, MV Buses 8.2 8.3.1-1 Onsite AC Electrical Power System 8.3 8.3.2-1 Class 1E DC Power System 8.3 8.3.2-2 Non-Class 1E DC Power System 8.3 8.3.2-3 Class 1E 120 VAC Instrumentation and Control Power System 8.3 8.3.2-4 Non-Class 1E 120 VAC Instrumentation and Control Power 8.3 System 1.7-5 Rev. 2

APR1400 DCD TIER 2 Table 1.7-2 (1 of 3)

APR1400 System Flow Diagrams DCD Tier 2 Figure No. Title Section 1.7-1 Flow Diagram Symbols and Legend 1.7 5.1.1-1 Reactor Coolant System Schematic Flow Diagram 5.1 5.1.2-1 Reactor Coolant System Flow Diagram 5.1 5.1.2-2 Reactor Coolant Pump Flow Diagram 5.1 5.1.2-3 Pressurizer and POSRV Flow Diagram 5.1 5.4.12-1 Reactor Coolant Gas Vent System Flow Diagram 5.4 6.2.2-1 Containment Spray System Flow Diagram 6.5 6.3.2-1 Safety Injection / Shutdown Cooling System Flow Diagram 6.3 6.4-1 Control Room Envelope Flow Diagram 6.4 6.4-2 Control Room HVAC System Flow Diagram (Normal Mode) 6.4 6.4-3 Control Room HVAC System Flow Diagram (Emergency Mode) 6.4 6.4-4 Control Room HVAC System Flow Diagram (Recirculation Mode) 6.4 6.8-3 In-Containment Water Storage System Flow Diagram 6.8 9.1.3-1 Spent Fuel Pool Cooling and Cleanup System Flow Diagram 9.1 9.2.1-1 Essential Service Water System Flow Diagram 9.2 9.2.2-1 Component Cooling Water System Flow Diagram 9.2 9.2.4-1 Domestic Water and Sanitary System Flow Diagram 9.2 9.2.5-1 Ultimate Heat Sink (Cooling Tower) Flow Diagram 9.2 9.2.6-1 Makeup Demineralizer System Flow Diagram 9.2 9.2.6-2 Condensate Storage System and Transfer System Flow Diagram 9.2 9.2.7-1 Essential Chilled Water System Flow Diagram 9.2 9.2.7-2 Plant Chilled Water System Flow Diagram 9.2 9.2.8-1 Turbine Generator Building Closed Cooling Water System Flow 9.2 Diagram 9.2.9-1 Turbine Generator Building Open Cooling System Flow Diagram 9.2 9.3.1-1 Instrument Air System Flow Diagram 9.3 9.3.1-2 Service Air System Flow Diagram 9.3 9.3.2-1 Process and Post-Accident Sampling System Flow Diagram 9.3 1.7-6 Rev. 2

APR1400 DCD TIER 2 Table 1.7-2 (2 of 3)

DCD Tier 2 Figure No. Title Section 9.3.2-2 Process Sampling System Flow Diagram 9.3 9.3.3-1 Radioactive Drain System Flow Diagram 9.3 9.3.4-1 Chemical and Volume Control System Flow Diagram 9.3 9.4.1-1 Control Room HVAC System Flow Diagram 9.4 9.4.2-1 Fuel Handling Area HVAC System Flow Diagram 9.4 9.4.3-1 Auxiliary Building Clean Area HVAC System Flow Diagram 9.4 9.4.4-1 Turbine Generator Building HVAC System Flow Diagram 9.4 9.4.5-1 Emergency Diesel Generator Area HVAC System Flow Diagram 9.4 9.4.5-2 Electrical and I&C Equipment Areas HVAC System Flow Diagram 9.4 9.4.5-3 Auxiliary Building Controlled Area HVAC System Flow Diagram 9.4 9.4.6-1 Reactor Containment Building HVAC System Flow Diagram 9.4 9.4.6-2 Reactor Containment Building Purge System Flow Diagram 9.4 9.4.7-1 Compound Building HVAC System Flow Diagram 9.4 9.5.1-1 Fire Protection System Flow Diagram 9.5 9.5.4-1 Diesel Fuel Oil Transfer System Flow Diagram 9.5 9.5.5-1 Emergency Diesel Engine Cooling Water System Flow Diagram 9.5 9.5.6-1 Emergency Diesel Engine Starting Air System Flow Diagram 9.5 9.5.7-1 Emergency Diesel Engine Lube Oil System Flow Diagram 9.5 9.5.8-1 Emergency Diesel Engine Air Intake and Exhaust System Flow Diagram 9.5 9.5.9-1 Gas Turbine Generator Facility Flow Diagram 9.5 10.1-1 Heat Balance Diagram 10.1 10.1-2 Overall System Flow Diagram 10.1 10.3.2-1 Main Steam Systems Flow Diagram 10.3 10.3.2-2 Turbine System Flow Diagram 10.3 10.4.2-1 Condenser Vacuum System Flow Diagram 10.4 10.4.3-1 Turbine Steam Seal System Flow Diagram 10.4 10.4.5-1 Circulating Water System Flow Diagram 10.4 10.4.6-1 Condensate Polishing System Flow Diagram 10.4 10.4.7-1 Condensate and Feedwater System Flow Diagram 10.4 1.7-7 Rev. 2

APR1400 DCD TIER 2 Table 1.7-2 (3 of 3)

DCD Tier 2 Figure No. Title Section 10.4.8-1 Steam Generator Blowdown System Flow Diagram 10.4 10.4.9-1 Auxiliary Feedwater System Flow Diagram 10.4 10.4.10-1 Auxiliary Steam System Flow Diagram 10.2 11.2-1 Liquid Radwaste System Flow Diagram 11.2 11.3-1 Gaseous Radwaste System Flow Diagram 11.3 11.4-1 Solid Radwaste System Flow Diagram 11.4 1.7-8 Rev. 2

APR1400 DCD TIER 2 EQUIPMENT SYMBOLS KNIFE GATE VALVE OTHER OPERATORS PIPE LINE COMPONENT SYMBOLS PRIMARY CONTAINMENT DOUBLE ACTING LETTER PREFIX CODE FUME SCRUBBER DOUBLE BLADE DAMPER (AIR TO OPEN AND AIR TO CLOSE) WELDED CAP TURBINE PC PENETRATION OF V SCREWED CAP V : VENT 025 PRIMARY CONTAINMENT VALVE POSITION STATUS (TYPICAL FOR ALL VALVES) D : DRAIN SINGLE FLOW DOUBLE FLOW E H

ELECTRO-HYDRAULIC ACTUATORS PENETRATION NUMBER BLIND FLANGE STEAM GENERATOR LO LETTER PREFIX CODE :

PRESSURIZED GAS BOTTLE LO : LOCKED OPEN IN LOCAL HOSE CONNECTION (COUPLER) PC : PIPING HAND ACTUATOR IC : INSTRUMENTATION PIPING (NOT SHOWN UNLESS AS INDICATED HOSE CONNECTION LC LC : LOCKED CLOSED IN LOCAL ABOVE OR WHEN USED AS (QUICK DISCONNECT)

AN EXTENSION SHAFT THRU SHIELD BREATHER CAP WALL OR FOR FLOORSTAND)

REACTOR VESSEL LO INSIDE SHIELDED AREA LO : LOCKED OPEN ON MCB AND/OR CONCENTRIC REDUCER OR ENLARGER REMOTE SHUTDOWN CONSOLE PENETRATION OF SHIELDED WALL MANUAL FLOW CONTROL ECCENTRIC REDUCER OR ENLARGER BALL STRAINER HORIZONTAL CENTRIFUGAL PUMP LC IF : INSULATED FLANGE LC : LOCKED CLOSED ON MCB AND/OR DU DU : DIELECTRIC UNION (SHOWN) VALVE REACH ROD REMOTE SHUTDOWN CONSOLE OPERATOR FAILURE MODE ABBREVIATION FLANGED CONNECTION LT LT : LOCKED THROTTLED FO : FAIL OPEN FO PLUGGED TAP ROOF BUILDING ROOF VERTICAL PUMP FC : FAIL CLOSED RESIN FILL CHUTE FC SELF-ACTUATED VALVES IDENTIFICATION OF CONNECTION CANNED WET PIT SUMP AT CONDENSER FL : FAIL LOCKED IN PLACE REMOVAL SPOOL FL (POSITION DOES NOT CHANGE)

VALVE IDENT. NUMBER (TYPICAL) PROCESS DATA POINT IDENTIFICATION POSITIVE DISPLACEMENT PUMP SET @_ _ _ _ PSIG ROOF DRAIN RECIPROCATING ROTARY TANDEM LOUVER DAMPER ANGLE SAFETY MISCELLANEOUS PIPING SPECIALTY SYMBOLS FLOOR DRAIN OR RELIEF DE (OR ED)

TRAP PLUGGED FLOOR DRAIN SAFETY CLASSIFICATION VACUUM PUMP ST : STEAM TRAP (SHOWN) OR SURFACE COLLECTION UNIT SET @_ _ _ _ PSIG ST AT : AIR TRAP DT : DRAIN TRAP AMMONIA VAPORIZER STRAIGHT SAFETY RT : RESIN TRAP EQUIPMENT DRAIN OR RELIEF DE (OR ED) QUALITY GROUP/SAFETY CLASS Y-TYPE STRAINER SEISMIC CATEGORY SET @_ _ _ _ PSIG HAND PUMP TEMPORARY STARTUP STRAINER INSTRUMENT DRAIN D CI PRESSURE REDUCING REGULATOR DE (OR ED)

SELF-CONTAINED SWING PANEL CIRCULAR OR HAMMER BLIND PLATE TYPE HEAT EXCHANGER PROTECTIVE EQUIPMENT DRAIN SPECTACLE BLIND DE SET @_ _ _ _ IN (H20,HG)

NOTE : THE QUALITY GROUP IDENTIFIERS SHALL ALSO SIMPLEX BASKET STRAINER LOOP SEAL REPRESENT SAFETY CLASS AS INDICATED BELOW VACUUM RELIEF U-TUBE HEAT EXCHANGER IRWST SUMP STRAINER DUPLEX BASKET STRAINER RUNNING TRAP QUALITY GROUP SAFETY CLASS SET @_ _ _ _ PSIG STRAIGHT TUBE A 1 P PILOT OPERATED MIXING TEE P TRAP HEAT EXCHANGER B 2 VALVE SYMBOLS SAFETY RELIEF C OR G 3 SINGLE-PASS TWO-PASS D OR E NNS BURSTING PLUG ATMOSPHERIC VENT PRESSURE VESSEL, VERTICAL LINE VALVE-BODIES (WITH SCREEN)

SET @_ _ _ _ PSIG OR HORIZONTAL TANK OPEN CLOSED DURING DURING OPERATOR SYMBOLS RUPTURE DISC FOR PRESSURE RELIEF ATMOSPHERIC VENT BUILDING TRANSITION (FOR USE WITH VALVE & DAMPER SYMBOL)

NORMAL NORMAL SET @_ _ _ _ PSIG (WITHOUT SCREEN)

DOME ROOF TYPE TANK OPERATION OPERATION VALVE TYPE TGB AB VALVE IDENTIFICATION RUPTURE DISC FOR VACUUM RELIEF THERMAL SLEEVE GATE V001A SET @_ _ _ _ PSIG INSTRUMENT IDENTIFICATION VALVE NO. GLOBE (ARROW INDICATES VALVE S NUMBER FOR CONTROL AIR & RESTRICTION ORIFICE CYLINDRICAL TYPE TANK BODY ORIENTATION. THIS ARROW INSTRUMENT AIR SIGNAL (SINGLE OR MULTI-STAGE) RELIEF VALVE VENT PIPE (DRUM TYPE) SHOW ONLY WHEN DIFFERENT FROM SOLENOID VALVES ONLY SLIP JOINT TGB : TURBINE GENERATOR BUILDING NORMAL SYSTEM FLOW DIRECTION OR REQUIRED IN BI-DIRECTION FLOW.) RESTRICTION DEVICE AB : AUXILIARY BUILDING CAP (CAPILLARY) RCB : REACTOR CONTAINMENT BUILDING BUTTERFLY VALVE IDENTIFICATION NUMBER FA : FUEL HANDLING AREA NO NC (TYPICAL FOR POWER OPERATED CAVITATING VENTURI CB : COMPOUND BUILDING CONE ROOF TANK BALL (2-WAY) VALVES EXCEPT AS NOTED FOR CERTAIN SOLENOID VALVES)

PIPE LINE SYMBOLS WT : FIRE PUMP AND WATER TREATMENT BUILDING CL : CHLORINATION BUILDING EXPANSION JOINT BALL VALVE (3-WAY) ST : SANITARY TREATMENT FACILITY D

UCT : UNDERGROUND COMMON TUNNEL D : DIAPHRAGM TANK (AS SHOWN) FLEXIBLE HOSE MAIN PROCESS FLOW IS : INTAKE STRUCTURE PLUG (2-WAY) MOTOR OPERATOR F : FLOATING TANK COVER EDGB : EMERGENCY DIESEL GENERATOR BUILDING FLEXIBLE CONNECTION PLUG VALVE (3-WAY)

SECONDARY PROCESS FLOW PLUG VALVE (4-WAY) M M PULSATION DAMPENER QUALITY GROUP ISOLATION FLAT TYPE TANK ANGLE LOOP SEAL DC AC MOTOR FA FLAME ARRESTOR DC MOTOR 1. AN ISOLATION DEVICE SUCH AS VALVE SHALL BE PROVIDED SLOPING LINE AT THE TRANSITION BOUNDARY BETWEEN TWO DIFFERENT ANGLE CHECK VALVE CONE BOTTOM TYPE TANK QUALITY GROUPS. THE ISOLATION DEVICE SHALL BE ASSIGNED OIL SEAL POT (DOWNWARD) UPWARD OR CYCLONE PNEUMATIC OPERATORS (SPRING RETURN) BY THE HIGHER QUALITY GROUP.

DIAPHRAGM FOR CONTROL AND ON/OFF VALVES HEAT TRACING (TYPICAL FOR OTHER VALVE TYPES) S AIR COMPRESSOR FG FLOW SIGHT GLASS (S : STEAM TRACED)

PACKLESS VALVE E RE : RECIPROCATING (SHOWN) B B (E : ELECTRICALLY TRACED)

B : BELLOWS SEAL (AS SHOWN)

RE RO : ROTARY M : METALLIC DIAPHRAGM DIAPHRAGM CYLINDER FLOW STRAIGHTENING VANE SYSTEM TRANSITION SC : SCREW OPERATOR OPERATOR CF : CENTRIFUGAL NEEDLE (TYPICAL) (TYPICAL) GUARDED LINE F FILTER AFTER COOLER FOOT VALVE 1. SYSTEM TRANSITION SHALL BE DENOTED AND REPRESNTED D DESICCANT FILTER SIMILAR TO BUILDING TRANSITION.

BURIED PIPE GLOBE GATE NON-RETURN VALVE VALVE VALVE GENERIC BOX EQUIPMENT (AIR TO CLOSE, i.e. FAIL OPEN)

ROOF EXHAUST HOOD TRENCH PIPE TRENCH RAM VALVE SPRAY NOZZLE OR SPARGER GENERIC CIRCLE EQUIPMENT OPERATOR UNLOADED OPERATOR LOADED (UPWARD OR DOWNWARD)

BACKFLOW PREVENTER (TYPICAL) (TYPICAL)

TUNNEL PIPE TUNNEL ELECTRIC HEATER CHECK IDENTIFY CONTROLLED HOPPER CLOSURE CHECK VALVES AND TILTING (AIR TO OPEN, i.e. FAIL CLOSED)

DISC CHECK VALVE BY NOTE M AIR SEPARATOR ASSEMBLY MISCELLANEOUS SYMBOLS STOP-CHECK SOLENOID OPERATORS MIXER NO NC (SPRING RETURN) FOR VALVES (TYPICAL FOR OTHER VALVE TYPES) SYSTEM ABBREVIATION OR DESCRIPTION.

SINGLE END SHUTOFF WITH CONDENSATE POT IF THIS SYMBOL INDICATES ANOTHER SHEET M ELECTRIC MOTOR QUICK DISCONNECT ON SAME SYSTEM, IT CAN BE OMITTED.

S S S S FLOW LIMITING VALVE MS LINE AND TUBING LINE CONTINUATION DEBRIS FILTER COMPRESSION TANK VENT FITTING SHEET 2 "A" ARROW HEAD (DIFFERENT FIGURE SHEET)

ENERGIZE TO CLOSE ENERGIZE TO OPEN PLUNGER VALVE (FAIL OPEN) (FAIL CLOSED) SHEET NO. THESE CAN BE BLANKED WHEN A RELATED E SEQUENCE LETTER FIGURE OR SEQUENCE IS NOT INCLUDED VAPORIZER WATER SEALED WS WS DELAY COIL IN DCD E : ELECTRIC VAPORIZER (SHOWN) (TYPICAL FOR OTHER A : AMBIENT VAPORIZER VALVE TYPES) LINE AND TUBING LINE CONTINUATION CLOSED PORT IN A S S DIAPHRAGM SEAL ARROW HEAD (SAME FIGURE SHEET)

MODE SHOWN STEM LEAKOFF LINE HYDROGEN IGNITER SEQUENCE LETTER (TYPICAL FOR OTHER (ON SAME FIGURE SHEET)

VALVE TYPES) SILENCER DE-ENERGIZED CONTROL SIGNAL CONTINUATION EJECTOR OR EDUCTOR QUICK OPENING GATE ARROW HEAD (SAME FIGURE SHEET)

S S DESUPERHEATER CONTROL SIGNAL CONTINUATION ARROW STEAM CONVERTER CLOSED PORT IN THREE-WAY MODE SHOWN HEAD (C&ID OR DIFFERENT FIGURE SHEET)

CLOSED PORT IN DESUPERHEATER MODE SHOWN SAMPLE COOLER FOUR-WAY PIPE AND TUBING LINE CONTINUATION ENERGIZED (DIFFERENT FIGURE SHEET)

Figure 1.7-1 Flow Diagram Symbols and Legend (1 of 4) 1.7-9 Rev. 2

APR1400 DCD TIER 2 HVAC EQUIPMENT / DAMPERS SYMBOLS HVAC EQUIPMENT / DAMPERS SYMBOLS FIRE PROTECTION SYMBOLS

• PIPE LINE INSTRUMENT SYMBOLS FE FE CENTRIFUGAL FAN OR BLOWER OUTDOOR FIRE HYDRANT INSTRUMENT IDENT./NUMBER (TYPICAL)

DUCTED AIR FLOW VANE AXIAL FAN FLOW NOZZLE VENTURI TUBE NON-DUCTED AIR FLOW OUTDOOR FIRE HYDRANT WITH HOSE HOUSE FE METERING ORIFICE PLATE HUMIDIFIER NO NORMALLY OPEN ANGLE VALVE OF INDOOR FO FAIL OPEN FIRE HYDRANT FL FAIL LOCKED TYPICAL NC NORMALLY CLOSED UNCLASSIFIED FLOW ELEMENT FE FC FAIL CLOSED ALARM CHECK VALVE (MAGNETIC, TARGET, AIR FOIL, HC HEATING COIL(HC) ULTRASONIC, LAMINAR, etc. )

EL : ELECTRIC (SHOWN)

HW : HOT WATER POST INDICATOR VALVE S  : STEAM EL S F( ) ROTAMETER HYDRANT ISOLATION GATE VALVE SMOKE DAMPER CC COOLING COIL (CC) FE FE CW : CHILLED WATER (SHOWN) DRY VALVE CCW : COMPONENT COOLING WATER PITOT TUBE CW DX : DIRECT EXPANSION DRY VALVE WITH QUICK-OPENING DEVICE (ACCELERATOR OR EXHAUSTER) SINGLE PORT AVERAGING PREFILTER CLEAN AGENT, CO2, OR GAS FE SPRAY NOZZLE TURBINE OR PROPELLER TYPE METER D DELUGE VALVE HIGH EFFICIENCY PARTICULATE AIR TEST CONNECTIONS (HEPA) FILTER SPRINKLER OR WATER AIR FILTER-PRV SET (SELF RELIEVING)

SPRAY NOZZLE P

PRESSURE TEST CONNECTION PROCESS RADIATION MONITOR AUTO DRIP VALVE RE CARBON ADSORBER (ON-LINE TYPE)

AD G CONSTANT HEAD CHAMBER INDOOR FIRE HYDRANT WITH HOSE (RESERVOIR)

GRAB SAMPLE TEST CONECTION HOUSE (WET TYPE)

POSTFILTER INDOOR FIRE HYDRANT WITH HOSE RES CONSTANT HEAD CHAMBER HOUSE (DRY TYPE) (RESERVOIR WITH TEMP EQUAL COL)

T TEMPERATURE TEST CONNECTION (WHEN THERMOWELL IS NOT USED)

MOISTURE SEPARATOR CO2 HOSE REEL STATION TW THERMOWELL FOR TEST F

FLOW TEST CONNECTION FOAM HOSE STATION TEW THERMOWELL ROOF SUPPLY FAN FOAM CHAMBER X TE TEMPERATURE ELEMENT SPECIAL TEST CONNECTION (DESCRIBE) FOAM PROPOTIONER ROOF VENTILATOR WITH PARALLEL BLADE DAMPER

• NOTE : FOR SPRAY NOZZLES OR FS SPRINKLER HEADS. USE THE FLOW SWITCH X PIPING SPECIALLTY SYMBOL FOR SPRAY NOZZLE SPECIAL TEST CONNECTION (DESCRIBE) AE ANALYZER CELL (INLINE)

EVAPORATOR CONDENSER CHILLER SIAMESE FIRE HOUSE CONNECTION AX ASME TEST POINT ONLY(TYP. )

SAMPLE PROBE SUPPRESSION GAS BOTTLE CC CUBICLE COOLER WITH CENTRIFUGAL FAN CW

• MANUAL PULL STATION ALARM BOTH ASME AND NORMAL TEST POINT (TYP. ) EQUIPMENT NUMBER

• REFER TO ABOVE TEST PARAMETER :

P P, G, T, etc. . T TEST BOX PP 01 A-A1 SUBCOMPONENT NO.

NORMALLY OPEN DAMPER (CABLE ROUTING ONLY)*

(PNEUMATIC OPERATED) SPEAKER / HORN MULTIPLE CHARACTER (INDICATES SAFETY TRAIN)*

EQUIPMENT SERIAL NO.

EQUIPMENT TYPE (NOTE 1)

P BELL / GONG ABBREVIATIONS

• OPTIONAL NORMALLY CLOSED DAMPER (PNEUMATIC OPERATED) FIRE SYSTEM WMA WATER MOTOR ALARM HVAC ONLY EQUIPMENT TYPE (NOTE 1)

V V V OA  : OUTSIDE AIR CARBON ADSORBER DELUGE SYSTEMS E NOZZLES (HVAC) AC = ACCUMULATOR H ATM  : ATMOSPHERE AHU  : AIR HANDLING UNIT AP = AIR PREHEATER, STEAM AIR HEATER ACU  : AIR CLEANING UNIT BC = BATTERY CHARGER (RECTIFIER)

NORMALLY OPEN DAMPER CC  : CUBICLE COOLER CD = CONDENSER (ELECTRO-HYDRAULIC OPERATED) PACU  : PACKAGED AIR CONDITIONING UNIT CH = CHILLER SR  : SAFETY RELATED CM = COMPRESSOR NSR  : NON-SAFETY RELATED DD = DEMINERALIZER DW = DEHYDRATOR E EC = ECONOMIZER H ABBREVIATIONS FT = FILTER/ADSORBER/DUST COLLECTOR/MOISTURE SEPARATOR HE = HEAT EXCHANGER NORMALLY CLOSED DAMPER PA = I&C EQUIPMENT ROOM PANEL (CABINET)

(ELECTRO-HYDRAULIC OPERATED) PM = MAIN CONTROL ROOM CONSOLE AND PANEL FIRE PROTECTION ONLY RU = REMOTE SHUTDOWN ROOM CONSOLE RW = RADWASTE CONTROL ROOM CONSOLE LFCP  : LOCAL FIRE CONTROL PANEL LX = LOOP CONTROLLER CABINET SIC  : SIGNAL INTERFACE CABINET GX = GROUP CONTROLLER CABINET TC  : TRANSPONDER CABINET LDP = LARGE DISPLAY PANEL FH(S)  : FIRE HYDRANT PM05 = SAFETY CONSOLE FS  : FIRE SUPRESSION SYSTEM AWSS  : AUTO WET SPRINKLER SUPRESSION PP = PUMP AWPSS  : AUTO WET PREACTION SPRINKLER SUPRESSION RV = REACTOR VESSEL PARALLEL BLADE DAMPER SG = STEAM GENERATOR TA = TURBINE TK = TANK LP = LOCAL CONTROL PANEL MX = MULTIPLEXING CABINET Figure 1.7-1 Flow Diagram Symbols and Legend (2 of 4) 1.7-10 Rev. 2

APR1400 DCD TIER 2 INSTRUMENT & CONTROL DEVICE SYMBOLS CONTROL & INSTRUMENTATION FUNCTION SYMBOLS & CODES INSTRUMENT CLARIFYING SYMBOL LIST

• • SHARED HUMAN-SYSTEM INTERFACE (HSI)

(HAND SWITCH, INDICATOR, CONTROL, ALARM, RECORDER)

• • • WITHIN POWER BLOCK (1) FUNCTION SYMBOLS FUNCTION SYMBOL CODES CLARIFYING SYMBOL LIST A ANALOG SIGNAL ADD ADD

• • SHARED HSI (HAND SWITCH, INDICATOR,CONTROL, ALARM, AVG AVERAGE RECORDER)

OUT OF POWER BLOCK (1) BI BIAS C CONDUCTIVITY VARIABLE (MEASURED) CC CATION CONDUCTIVITY CL- CHLORIDE LOCAL FUNCTION OF VARIABLE CHLORINE CL2

• • ELECTRICAL SIGNAL BETWEEN LOCAL & MAIN CO2 CARBON DIOXIDE MAIN CONTROL ROOM PANEL MOUNTED DEVICE (FIXED) CONTROL ROOM / I & C EQUIPMENT PTIS ROOM FUNCTIONS INSTRUMENT NUMBER D DIGITAL 001 SUBTRACT 1PM05 CONNECTION FOR THE PROCESS VARIABLE DIFF

• PANEL NUMBER (TYPICAL) (TYPICAL) XCIR O2 CLARIFYING SYMBOL PER ABOVE SYMBOL LIST (IF REQUIRED)

DIR DIRECT ACTING E/I CONVERTER (VOLTAGE TO CURRENT)

FC FAIL CLOSED F/H FOLLOW / HOLD FI FAIL INTERMEDIATE CONTROL DEVICE AND / OR CCS CABINET MOUNTED DEVICE FUNCTION IN MAIN CONTROL ROOM AREA FL FAIL LOCK INSTALLED WITHIN POWER BLOCK FO FAIL OPEN LOCAL FUNCTION VARIABLE IN MAIN CONTROL ROOM AREA 1PA01 H2 HYDROGEN (USED WHEN NO LOCAL VARIABLE EXISTS)

• PANEL NUMBER (TYPICAL) HE HELIUM HHL HIGH HIGH LIMIT HL HIGH LIMIT HP HIGH PASS CONTROL DEVICE AND / OR CCS CABINET MOUNTED DEVICE MULTIPLE VARIABLE WHEN APPLICABLE I/E CONVERTER (CURRENT TO VOLTAGE)

INSTALLED OUT OF POWER BLOCK INT INTEGRATOR DASH USED WHEN MULTIPLE VARIABLES EXIST I/P CONVERTER (ELECTRIC CURRENT TO PNEUMATIC)

ISO ISOLATOR PF-TIS LL LOW LIMIT 001 ALTERNATE WHEN REQUIRED FOR LLL LOW LOW LIMIT LOCALLY MOUNTED DEVICE NOT ON MULTIPLE VARIABLES LP LOW PASS INSTRUMENT PANEL OR ASSEMBLY CIR MANUAL M

MAX MAXIMUM M/A MANUAL / AUTOMATIC MIN MINIMUM LOCAL INSTRUMENT PANEL N2 NITROGEN MOUNTED DEVICE NA SODIUM NAK SODIUM POTASSIUM 1LP01 N 2H 2 HYDRAZINE

• PANEL NUMBER (TYPICAL) O2 OXYGEN pH pH ANALYSIS RES RESET REN REVERSE ACTING MULTIPLE DEVICES IN ONE ASSEMBLY SC SPECIFIC CONDUCTIVITY (i.e. TEMPERATURE ELEMENT IN THERMOWELL) SI SILICA SM SMOKE OR IONIZATION PLANT UNIT NUMBER NOT REQUIRED IF SAME SQRT SQUARE ROOT AS DRAWING PLANT UNIT NUMBER TURB TURBIDITY VOL VOLUME THE INDICATED POWER SEGREGATION REPRESENTS X MULTIPLY THE POWER SEGREGATION OF COMPONENT REF REFRIGERANT BEING CONTROLLED BY THE SHARED CONTROL. DC DEDICATED CONTROL IPS INFORMATION PROCESSING SYSTEM TYPE OF HSI DEVICE SUCH AS QIAS-N, QIAS-P QIAS-N QUALIFIED INDICATION AND ALARM SYSTEM - NON-SAFETY OR LDP IS PRESENTED AS SUFFIX. QIAS-P QUALIFIED INDICATION AND ALARM SYSTEM - P HOWEVER, THE TYPE OF HSI DEVICE IS NOT SC SOFT CONTROL REQUIRED IF THE SYMBOL IS USED FOR THE SHARED CONTROL.

THE SHARED SYMBOL WITHOUT SUFFIX MEANS THIS INSTRUMENT TYPE CODES SYSTEM CLARIFYING SYMBOL LIST PARAMETER IS ONLY REPRESENTED ON THE INFORMATION FPD. THE QIAS-N MEANS THIS PARAMETER IS REPRESENTED ON BOTH INFORMATION FPD AND QIAS-N FPD.

THE LDP MEANS THIS PARAMETER IS REPRESENTED ON THE CLARIFYING SYMBOL LIST INFORMATION FPD, QIAS-N AND LDP.

FOR A MAN SWITCH, ONLY THE COMPONENT STATUS IS AF AUXILIARY FEEDWATER SYSTEM DISPLAYED ON QIAS-N AND LDP. FIRST LETTER SECOND OR SUCCEEDING LETTERS AS AUXILIARY STEAM VARIABLE INSTRUMENT FUNCTION CA CONDENSER VACUUM SYSTEM (1) SWITCH YARD IS EXCLUDED FROM POWER BLOCK AREA. CC COMPONENT COOLING WATER SYSTEM (MEASURED OR INITIATING) CD CONDENSATE SYSTEM CM CONTAINMENT MONITORING SYSTEM A ANALYSIS A ALARM CP CONDENSATE POLISHING SYSTEM B BURNER AC ALARM CLOSED CS CONTAINMENT SPRAY SYSTEM C CONDUCTIVITY AH ALARM HIGH COMPUTER INPUT - OUTPUT SIGNAL NUMBER D DENSITY AHH ALARM HIGH HIGH CT CONDENSATE STORAGE AND TRANSFER SYSTEM CHEMICAL AND VOLUME CONTROL SYSTEM CV E VOLTAGE AHL ALARM HIGH OR LOW CW CIRCULATING WATER SYSTEM F FLOW AL ALARM LOW DA GAS TURBINE GENERATOR FACILITY SYSTEM H HAND (MANUAL) ALL ALARM LOW LOW DE RADIOACTIVE DRAIN SYSTEM ALL OF INPUT POINTS TO CCS ARE REGARED AS THE COMPUTER I CURRENT (ELEC.) AO ALARM OPEN DG EMERGENCY DIESEL GENERATOR SYSTEM INPUT DATA FOR IPS OR QIAS-N. J POWER (ELEC.) C CONTROLLER (BLIND) DO DIESEL FUEL OIL TRANSFER SYSTEM K TIME E PRIMARY ELEMENT ED NON-RADIOACTIVE EQUIPMENT VENT & DRAIN SYSTEM L LEVEL(LIQUID OR SOLID) EW PRIMARY ELEMENT WELL FC SPENT FUEL POOL COOLING AND CLEANUP SYSTEM M MOISTURE (HUMIDITY) G GLASS FP FIRE PROTECTION SYSTEM N NEUTRON FLUX I INDICATOR LOCAL PANEL INPUT SYMBOL P PRESSURE OR VACUUM IC INDICATING CONTROLLER FW FEEDWATER SYSTEM GW GASEOUS RADWASTE MANAGEMENT SYSTEM PD DIFFERENTIAL PRESSURE IK INDICATING CONTROL STATION (MANUAL/AUTO) IA INSTRUMENT AIR SYSTEM R RADIOACTIVITY IR INDICATING RECORDER IW IN-CONTAINMENT WATER STORAGE SYSTEM INPUT SIGNAL FROM LOCAL S SPEED OR FREQUENCY IS INDICATING SWITCH MS MAIN STEAM SYSTEM TO CCS PANEL TO COMPUTER T TEMPERATURE IT INDICATING TRANSMITTER PR PROCESS RADIATION MONITORING SYSTEM THROUGH CCS U MULTIVARIABLE IY INDICATING RELAY PS PROCESS SAMPLING SYSTEM V VIBRATION JR MULTIPOINT RECORDER (MORE THAN 3 PENS) RC REACTOR COOLANT SYSTEM W WEIGHT OR FORCE K CONTROL STATION-VARIABLE TYPE RG REACTOR COOLANT GAS VENT SYSTEM X SPECIAL (REFER TO CLARIFYING SYMBOL LIST) (MANUAL/AUTO) WITHOUT CONTROLLER SA SERVICE AIR SYSTEM TO QIAS-P INPUT SIGNAL TO QIAS-P COMPUTER Y STRAIN L LIGHT (PILOT) SD STEAM GENERATOR BLOWDOWN SYSTEM Z POSITION Q INTEGRATOR SI SAFETY INJECTION / SHUTDOWN COOLING SYSTEM R RECORDER SX ESSENTIAL SERVICE WATER SYSTEM RC RECORDER CONTROLLER VB COMPOUND BUILDING HVAC SYSTEM S SWITCH VC CONTROL ROOM HVAC SYSTEM TO DIS INPUT SIGNAL TO DIS COMPUTER T TRANSMITTER VD EMERGENCY DIESEL GENERATOR AREA HVAC SYSTEM U MULTIFUNCTION VE ELECTRICAL AND I&C EQUIPMENT AREAS HVAC SYSTEM W WELL VF FUEL HANDLING AREA HVAC SYSTEM X UNCLASSIFIED VK AUXILIARY BUILDING CONTROLLED AREA HVAC SYSTEM Y RELAY, SIGNAL CONVERTER, COMPUTING SYSTEM LEVEL ACTUATION SYMBOL SOLENOID, ETC.

VO AUXILIARY BUILDING CLEAN AREA HVAC SYSTEM VP REACTOR CONTAINMENT BUILDING HVAC SYSTEM Z POWER POSITIONER / CONTROL DRIVE VQ REACTOR CONTAINMENT BUILDING PURGE SYSTEM (EXCEPT VALVE MOUNTED) VT TURBINE GENERATOR BUILDING HVAC SYSTEM ESF-SIAS SYSTEM LEVEL WD DOMESTIC WATER SYSTEM START ESF ACTUATION SIGNAL WH TURBINE GENERATOR BUILDING OPEN COOLING SYSTEM WI PLANT CHILLED WATER SYSTEM WO ESSENTIAL CHILLED WATER SYSTEM WT TURBINE GENERATOR BUILDING COMPONENT COOLING SYSTEM DMA-SIAS DIVERSE MANUAL WV LIQUID RADWASTE MANAGEMENT SYSTEM START ACTUATION SIGNAL WX SOLID RADWASTE MANAGEMENT SYSTEM Figure 1.7-1 Flow Diagram Symbols and Legend (3 of 4) 1.7-11 Rev. 2

APR1400 DCD TIER 2 TYPICAL INSTRUMENT & CONTROL SYMBOL EXAMPLES INSTRUMENT SENSING PIPE LINE SYMBOLS VALNE OPERATOR CONTROLS

( TYPICAL FOR OTHER CONTROL DEVICES )

S DOUBLE ACTING CYLINDER OPERATOR MULTIPLE CONTROL SIGNALS FOR PNEUMATIC OPERATORS IA AIR TO OPEN AND AIR TO CLOSE (ENERGIZE TO OPEN-ENERGIZE TO CLOSE)

S INSTRUMENT IDENTIFICATION NUMBER CAPILLARY TUBING (FILLED SYSTEM )

LINE NOT CONNECTED S S DOUBLE ACTING CYLINDER OPERATOR AIR TO OPEN AND AIR TO CLOSE IA IA (LOSS OF ELECTRICAL POWER CLOSES VALVE)

INSTRUMENT AIR SUPPLY SYSTEM CODE ( TYPICAL)

DIAPHRAGM OPERATOR (TYPICAL)

IA FILLED INSTRUMENT SENSING LINE SEAL SPRING - RETURN OPERATOR S (ENERGIZE TO OPEN AND DE-ENERGIZE TO CLOSE VALVE ON ELECTRICAL OR AIR FAILURE)

SOLENOID VALVE FURNISHED MOUNTED WITH OPERATOR S

SINGLE CONTROL SIGNALS FOR PNEUMATIC OPERATORS INSTRUMENT & CONTROL SIGNAL SYMBOLS PNEUMATIC ANALOG SIGNAL LINE LINE NOT CONNECTED LINE CONNECTED VALVE / DAMPER POSITION SIGNALS ELECTRICAL ANALOG OR DIGITAL SIGNAL HS OPEN & CLOSED POSITION LIGHTS LINE NOT CONNECTED S

INCLUDED WITH HAND SWITCH (EXCEPT M.O.TYPE)

LINE CONNECTED OPTIONAL POSITION FROM STEM-MOUNTED DEVICES FURNISHED INTEGRAL WITH VALVE / DAMPER (EXCEPT M.O. TYPE)

HYDRAULIC SIGNAL LINE STEM MOUNTED POSITION DEVICE PER CONTROL VALVE / DAMPER DATA SHEET LINE NOT CONNECTED LINE CONNECTED OPEN & CLOSE POSITION LIGHTS ZL MOUNTED REMOTE FROM HAND SWITCH OR WHEN NO HAND SWITCH INCLUDED ON-OFF CONTROLLING DEVICE (TYPICAL)

HS S OPEN & CLOSE POSITION LIGHTS (OPERATED FROM MOTOR GEAR TRAIN)

IA DIAPHRAGM OPERATOR (TYPICAL)

INCLUDED WITH HAND SWITCH FOR M

MOTOR OPERATED VALVES AND DAMPERS NOTE : USED ONLY WHEN SOLENOID NOT (STATUS LIGHTS MAY BE SHOWN FOR MOUNTED ON OPERATOR MOV ON THE SAME BASIS AS ABOVE)

INSTRUMENT AIR SUPPLY SIGNAL SYSTEM CODE (TYPICAL)

Figure 1.7-1 Flow Diagram Symbols and Legend (4 of 4) 1.7-12 Rev. 2

APR1400 DCD TIER 2 1.8 Interfaces with Standard Designs The APR1400 design includes an essentially complete nuclear plant, but does not include the structures, systems, and components (SSCs) that are part of the site-specific designs.

Section 1.8 provides a list of the interface requirements for SSCs within the scope of the APR1400 that are required by 10 CFR 52.47(a) (Reference 1).

A standard site plot of the APR1400 is provided in Figure 1.2-1. The plot shows the scope of the design certification application. To provide reasonable assurance that the site-specific SSCs are compatible with the APR1400 design, interface requirements for site-specific SSCs to be satisfied by the combined license (COL) applicant are provided in the sections referenced in Table 1.8-1. In general, interface requirements for COL applicant-supplied SSCs that relate to a specific mechanical or electrical system are addressed in the appropriate chapter. The conceptual design information (CDI) for those portions of the plant for which the application does not seek certification is indicated by double brackets in the text and tables and cloud marks in the figures. Safety-significant interface requirements for site-specific SSCs are specified in Tier 1 of the DCD.

Table 1.8-1 is an index of all sections of this document that contain interface requirements.

The COL applicant is to describe how the site-specific interface requirements are met (COL 1.8(1)).

Site characteristic assumptions on which the APR1400 design is based are presented in Chapter 2. The site characteristics are to be compatible with the APR1400 design envelopes but are not considered interface requirements as used in 10 CFR 52.47(a).

Table 1.8-2 presents the list of all COL information items. The COL applicant is to identify how each COL information item is addressed (COL 1.8(2)).

1.8.1 Combined License Information COL 1.8(1) The COL applicant is to describe how site-specific interface requirements are met.

COL 1.8(2) The COL applicant is to identify how each COL information item is addressed.

1.8-1 Rev. 2

APR1400 DCD TIER 2 1.8.2 References

1. 10 CFR 52.47(a), Contents of Applications; Technical Information.

1.8-2 Rev. 2

APR1400 DCD TIER 2 Table 1.8-1 (1 of 2)

Index of System, Structure, or Component Interface Requirements for APR1400 Interface DCD Tier 2 System, Structure, or Component Type Section Structures Switchyard COL 8.2 Emergency operations facility COL 13.3.3.2 Ultimate heat sink, including ESWS intake/discharge CDI 9.2.5.2 Domestic water and sanitary system structure CDI 9.2.4 Circulating water pump house CDI 10.4.5.2 Normal plant heat sink, including CW system intake/discharge CDI 10.4.5 Seismic Category II buildings described in Table 3.2-1 COL 3.7.5 Component cooling water heat exchanger building COL 3.7.5 3.8.6 Essential service water building COL 3.7.5 3.8.6 Systems Offsite power system, including switchyard COL 8.2 Domestic water and sanitary systems, including sanitary water CDI 9.2.4 treatment facility Security system COL 13.6.1 Communication system (offsite) COL 9.5.2.2.2 UHS CDI 9.2.5 1.8-3 Rev. 2

APR1400 DCD TIER 2 Table 1.8-1 (2 of 2)

Interface DCD Tier 2 Structure, System or Component Type Section Components UHS cooling tower CDI 9.2.5.2.2.1 UHS piping, valves, and fittings CDI 9.2.5.2.2.2 UHS cooling tower basin CDI 9.2.5.2.2.3 UHS cooling tower basin screens CDI 9.2.5.2.2.4 Circulating water pumps CDI 10.4.5.2.1 Cooling tower CDI 10.4.5.2.3 Cooling tower basin CDI 10.4.5.2.3 Cooling tower basin screen CDI 10.4.5.2.3 Cooling tower makeup and blowdown pump CDI 10.4.5.2.3 ESW blowdown piping CDI 9.2.1.2.1 Condenser vacuum pressure of a high pressure alarm and CDI 10.4.1.5, turbine trip 10.4.2.2.2 Cooling tower chemical injection system CDI 10.4.5.2.3 Fuel oil storage tanks, fuel oil day tanks, and fuel oil transfer CDI Table 9.5.4-1 pumps Cation-bed ion exchanger vessels CDI Table 10.4.6-1 Mixed-bed ion exchanger vessels CDI Table 10.4.6-1 Spent resin holding tanks CDI Table 10.4.6-1 Resin holding tank CDI Table 10.4.6-1 Resin mixing and holding tank CDI Table 10.4.6-1 Resin traps CDI Table 10.4.6-1 TBCCW pumps and heat exchangers CDI Table 9.2.8-1 1.8-4 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (1 of 39)

Combined License Information Items Item No. Description COL 1.1(1) The COL applicant that references the APR1400 is to identify the actual plant site location.

COL 1.1(2) The COL applicant that references the APR1400 is to provide estimated schedules for the completion of construction and the start of commercial operation.

COL 1.2(1) The COL applicant is to prepare a complete and detailed site plan.

COL 1.4(1) The COL applicant that references the APR1400 design certification is to identify major agents, contractors, and participants for the construction and operation of the nuclear power plant.

COL 1.8(1) The COL applicant is to describe how site-specific interface requirements are met.

COL 1.8(2) The COL applicant is to identify how each COL information item is addressed.

COL 1.9(1) The COL applicant is to provide an evaluation of the conformance with the regulatory criteria for the site-specific portions and operational aspects of the facility.

COL 2.0(1) The COL applicant is to demonstrate that the APR1400 design meets the requirements imposed by the site-specific parameters and conforms with all design commitments and acceptance criteria if the characteristics of the site fall outside the assumed site parameters in Table 2.0-1.

COL 2.1(1) The COL applicant is to provide site-specific information on the site location and description of the site, exclusion authority and control, and population distribution as stated in NRC RG 1.206.

COL 2.2(1) The COL applicant is to provide site-specific information on nearby industrial, transportation, and military facilities as required in NRC RG 1.206.

COL 2.2(2) The COL applicant is to identify the DBE caused by nearby industrial, transportation, and military facilities and determine its design parameters.

COL 2.3(1) The COL applicant is to provide site-specific information on meteorology including regional climatology, local meteorology, onsite meteorological measurement program, estimated short-term atmospheric dispersion for accident release, and long-term atmospheric dispersion estimates for routine release as addressed in NRC RG 1.206.

COL 2.3(2) The COL applicant is to perform the radiological consequence analysis and demonstrate that the related dose limits specified in 10 CFR 50.34 and 10 CFR Part 50 Appendix I are not exceeded, if the site-specific /Q values exceed the bounding values described in Tables 2.3-1 to 2.3-12.

COL 2.4(1) The COL applicant is to provide the site-specific hydrologic information on probable maximum precipitation (PMP), probable maximum flood (PMF) on streams and rivers, potential dam failures, probable maximum surge and seiche flooding, probable maximum tsunami hazards, ice effects, cooling water canals and reservoirs, channel diversions, flood protection requirements, low water considerations, ground water, potential accidental release of liquid effluents in ground and surface water, and Technical Specifications and emergency operation requirements in accordance with NRC RG 1.206, NRC RG 1.59, and NRC JLD-ISG-2012-06.

1.8-5 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (2 of 39)

Item No. Description COL 2.5(1) The COL applicant is to provide the site-specific information on geology, seismology, and geotechnical engineering as required in NRC RG 1.206. The COL applicant is to conduct the probabilistic seismic hazard analysis (PSHA) and develop the site-specific GMRS using the PSHA results as required in NRC RG 1.208.

COL 2.5(2) The COL applicant is to confirm that the foundation input response spectra (FIRS) of each seismic Category I structure are completely enveloped by the CSDRS-compatible free-field response motions at the bottom elevation of each seismic Category I structure for a site with the low-strain shear wave velocity greater than 304.8 m/s (1,000 ft/s) up to the structural foundation elevation. Alternately, the COL applicant is to confirm that the FIRS of each seismic Category I structure are completely enveloped by the CSDRS for a hard rock site with a low-strain shear wave velocity of supporting medium for each seismic Category I structure greater than 2,804 m/s (9,200 ft/s).

COL 2.5(3) The COL applicant is to confirm that (i) the requirement for the site-specific weight densities of subsurface soils is to be no less than 2,002.3 kg/m3 (125 lb/ft3); (ii) the profiles of site-specific soil properties are generally increasing with depth from the ground surface in a manner similar to the general profile shapes shown in Tables 3.7A-1 through 3.7A-9 and Figures 3.7A-3 through 3.7A-11; (iii) the site-specific soil profiles have no inverse condition, i.e., the soil properties of a deeper soil layer are less than the properties of the soil layer above it; and (iv) the site-specific best estimate (BE), lower bound (LB), and upper bound (UB) strain-compatible soil shear wave velocity profiles, including backfill, are consistent with one of the APR1400 generic site conditions S1 through S4 and S6 through S9 considered for the standard design as shown in Tables 3.7A-1 through 3.7A-9 and Figures 3.7A-3 through 3.7A-11. The lower bound and upper bound shear wave velocity profiles are obtained, as defined in SRP Section 3.7.2, from the mean properties plus or minus one standard deviation, maintaining the minimum variation of 1.5GBE for the upper bound range and GBE/1.5 for the lower bound range, where GBE denotes the best estimate low-strain shear modulus. The lower bound low-strain shear wave velocity for the site-specific soil profile is not to be less than 304.8 m/s (1,000 ft/s) up to the structural foundation elevation.

COL 2.5(4) The COL applicant is to confirm that the site-specific profile for a HRHF site is consistent with the soil profile provided in Table 3.7B-3 and the site-specific GMRS determined at the finished grade are completely enveloped by the APR1400 HRHF response spectra. In addition, the COL applicant is to confirm that the FIRS of the seismic Category I structures are completely enveloped by the HRHF-compatible free-field response motions at the bottom elevation of each seismic Category I structure.

COL 2.5(5) The COL applicant is to perform a site-specific seismic analysis to develop in-structure response spectra at key locations using the procedure described in Appendix 3.7A if COL 2.5(2) and COL 2.5(3) above are not met. The COL applicants site-specific strain-compatible properties are to be consistent with the assumptions used in the SSI analyses including embedment and extent of backfill, as described in Appendix 3.7A. The COL applicant is to confirm that the site-specific 5% damped in-structure response spectra so generated are enveloped by the corresponding 5% damped in-structure response spectra provided in Appendix 3.7A. If this requirement is not satisfied, the COL applicant is to calculate the site-specific member forces and compare them with CSDRS member forces at key locations to determine whether further site-specific seismic design is required.

1.8-6 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (3 of 39)

Item No. Description COL 2.5(6) The COL applicant is to perform a site-specific seismic response analysis using the procedure described in Appendix 3.7B and the EPRI White Paper, Seismic Screening of Components Sensitive to High Frequency Vibratory Motions, if COL 2.5(4) is not met.

The COL applicant is to develop the site specific in-structure response spectra and compare them with the corresponding HRHF-based in-structure response spectra to determine whether further structural integrity (including member forces) and functionality evaluations are required.

COL 2.5(7) The COL applicant is to perform an evaluation of the subsurface conditions within the standard plant structure footprint based on the geologic investigation in accordance with NRC RG 1.132.

COL 2.5(8) The COL applicant is to evaluate the potential future surface deformation of tectonic and non-tectonic origin.

COL 2.5(9) If the potential for future surface deformation exists, the COL applicant needs to demonstrate the potential effects of surface deformation are within the design basis of the facility.

COL 2.5(10) The COL applicant is to confirm that the soil angle of internal friction below the footprint of the seismic Category I structures at their excavation depth is a minimum of 35 degrees.

COL 2.5(11) The COL applicant is to confirm that the dynamic properties of structural fill granular and the compressive strength of lean concrete to be used in construction of the APR1400 seismic Category I structures satisfy the requirements of structural fill granular provided in Table 2.0-1 and the minimum compressive strength of 140 kg/cm2 (2,000 psi) for the lean concrete.

COL 2.5(12) The COL applicant is to evaluate the potential for liquefaction occurring at the site adjacent to and under seismic Category I structures in accordance with NRC RG 1.198. In addition, the COL applicant is to evaluate the liquefaction potential for those seismic Category II structure foundations that if failed, could degrade the function of a seismic Category I SSC to an unacceptable safety level.

COL 2.5(13) The COL applicant will evaluate the allowable bearing capacity of the subsurface based on the site-specific properties of the underlying materials, including appropriate laboratory test data to evaluate strength, and considering local site effects, such as fracture spacing, variability in properties, and evidence of shear zones. If the site-specific allowable bearing capacity is less than the maximum bearing demands specified in Table 2.0-1, a site-specific evaluation shall be performed by a COL applicant using the APR1400 basemat model and methodology described in Subsection 3.8.5.

COL 2.5(14) The COL applicant will verify whether the predicted settlements within each building and between buildings exceed the maximum differential settlement specified in Table 2.0-1 for site suitability determination. If the predicted settlement exceeds the maximum value in Table 2.0-1, a detailed site specific evaluation shall be performed by COL applicant to demonstrate acceptability. The COL applicant will also meet settlement criteria specified in COL 3.8(18) for construction sequence and post construction settlement limits.

COL 2.5(15) The COL applicant is to provide site-specific information about the static and dynamic stability of all natural and man-made soil and rock slopes including embankments and dams.

COL 3.2(1) The COL applicant is to identify the seismic classification of site-specific SSCs that should be designed to withstand the effects of the SSE.

1.8-7 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (4 of 39)

Item No. Description COL 3.2(2) The COL applicant is to identify the quality group classification of site-specific systems and components and their applicable codes and standards.

COL 3.2(3) The COL applicant is to provide the classification of structures, systems, and components for turbine generator building drain system.

COL 3.3(1) The COL applicant is to demonstrate that the site-specific design wind speed is bounded by the design wind speed of 64.8 m/s (145 mph).

COL 3.3(2) The COL applicant is to demonstrate that the site-specific seismic Category II structures adjacent to the seismic Category I structures are designed to meet the provisions described in Subsection 3.3.1.2.

COL 3.3(3) The COL applicant is to perform an analysis if the site-specific wind and tornado/hurricane characteristics are not bounded by the site parameter postulated for the certified design.

COL 3.3(4) The COL applicant is to provide reasonable assurance that site-specific structures and components not designed for the extreme wind loads do not impact either the function or integrity of adjacent seismic Category I SSCs.

COL 3.4(1) The COL applicant is to provide the site-specific design of plant grading and drainage.

COL 3.4(2) The COL applicant is to provide site-specific information on protection measures for the design-basis flood, such levees, seawells, flood walls, revetments, or breakwaters or site bulkheads pursuant to RG 1.102 as required in Subsection 2.4.10.

COL 3.4(3) The COL applicant is to establish procedures and programmatic controls to ensure the availability of the floor drainage.

COL 3.4(4) The COL applicant is to periodically inspect watertight doors and the penetration seals to ensure their functionality.

COL 3.4(5) The COL applicant is to provide flooding analysis with flood protection and mitigation features from internal flooding for the CCW Heat Exchanger Building and ESW Building.

COL 3.4(6) The COL applicant is to provide the site-specific flooding hazards from engineered features, such as water tank collapsing, water piping breaking, etc.

COL 3.4(7) The COL applicant is to confirm that the potential site-specific external flooding events are bounded by design-basis flood values or otherwise demonstrate that the design is acceptable.

COL 3.4(8) The COL applicant is to provide the site-specific dewatering system if the plant is built below the design basis flood level.

COL 3.4(9) The COL applicant is to describe the basis for the Probable Maximum Flood (PMF) to determine the maximum site-specific ground water elevation above the grade that may occur from tsunami or hurricane sources.

COL 3.4(10) The COL applicant is to identify any site-specific physical models that could be used to predict prototype performance of hydraulic structures and systems.

COL 3.5(1) The COL applicant is to provide the procedure for heavy load transfer to strictly limit the transfer route inside and outside containment during plant maintenance and repair periods.

COL 3.5(2) The COL applicant is to provide the procedures which ensure that equipment required during maintenance, should be removed from containment prior to operation, moved to a location where it is not a potential hazard to SSC important to safety, or seismically restrained.

1.8-8 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (5 of 39)

Item No. Description COL 3.5(3) The COL applicant is to perform an assessment of the orientation of the turbine generator of this and other unit(s) at multi-unit sites for the probability of missile generation using the evaluation of Subsection 3.5.1.3.2 to verify that essential SSCs are outside the low-trajectory turbine missile strike zone.

COL 3.5(4) The COL applicant is to evaluate site-specific hazards induced by external events that may produce more energetic missiles than tornado or hurricane missiles, and provide reasonable assurance that the seismic Category I structures are designed to withstand these loads.

COL 3.5(5) The COL applicant is to confirm that automobile missiles cannot be generated within a 0.5 mile radius of safety-related SSCs that would lead to impact higher than 10.06 m (33 ft) above plant grade.

COL 3.5(6) The COL applicant is to identify applicable tornado missile spectra and associate velocities for the compound building, and to evaluate the missile protection provided by the building.

COL 3.5(7) The COL applicant is to evaluate the potential for site proximity explosions and missiles due to train explosions (including rocket effects), truck explosions, ship or barge explosions, industrial facilities, pipeline explosions, or military facilities.

COL 3.5(8) The COL applicant is to provide justification for the site-specific aircraft hazard and an aircraft hazard analysis in accordance with the requirements of NRC RG 1.206.

COL 3.5(9) The COL applicant is to provide reasonable assurance that site-specific structures and components not designed for missile loads will not prevent safety-related SSCs from performing their safety function.

COL 3.6(1) The COL applicant is to identify the site-specific SSCs that are safety-related or required for safe shutdown that are located near high- and moderate-energy piping systems and that are susceptible to the consequences of piping failures.

COL 3.6(2) The COL applicant is to provide a list of site-specific high- and moderate-energy piping systems including layout drawings and protection features and the failure modes and effects analysis for safe shutdown due to the postulated HELBs.

COL 3.6(3) The COL applicant is to confirm that the bases for the LBB acceptance criteria are satisfied by the final as-built design and materials of the piping systems as site-specific evaluations, and is to provide the information including LBB evaluation report for the verification of LBB analyses.

COL 3.6(4) The COL applicant is to provide the procedure for initial filling and venting to avoid the known causes for water hammer in each piping system designed for LBB.

COL 3.6(5) The COL applicant is to provide the information on welding of Alloy 52/52M/152 concerning the residual stress and dilution effects of welds.

COL 3.7(1) The COL applicant is to demonstrate the applicability of soil degradation models used in site-specific site response analysis for the site conditions.

COL 3.7(2) The COL applicant is to compare the site-specific strain-compatible soil properties with generic soil properties in order to confirm that the site meets the generic soil profile used in the standard design.

1.8-9 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (6 of 39)

Item No. Description COL 3.7(3) The COL applicant is to provide the seismic design of the seismic Category I SSCs and seismic Category II structures that are not part of the APR 1400 standard plant design.

The seismic Category I and II structures are as follows:

a. Seismic Category I essential service water building

b. Seismic Category I component cooling water heat exchanger building

c. Seismic Category II turbine generator building

d. Seismic Category II compound building

e. Seismic Category II alternate alternating current gas turbine generator building COL 3.7(4) The COL applicant is to confirm that any site-specific non-seismic Category I structures are designed not to degrade the function of a seismic Category I SSC to an unacceptable safety level due to their structural failure or interaction. The COL applicant is to confirm that the calculated relative displacements do not exceed the gaps between seismic Category I and non-seismic Category I structures.

The COL applicant is to apply the site-specific FIRS as seismic input motions and to establish a site-specific soil profile as a supporting media for the seismic analysis of the seismic Category II structures. The COL applicant is to apply the same seismic analysis procedure as the seismic Category I structures to the seismic Category II structures. The COL applicant is to perform the structural design of the seismic Category II structures using the design codes described in Subsection 3.7.2.8 and Table 3.2-1. The COL applicant is to check the potential effects of sliding and uplift for the seismic Category II structures using the same approach applied in the stability check for the seismic Category I structures. The COL applicant is to evaluate the pressures on the below grade walls of the NI, resulting from site-specific SSSI effects between the NI and adjacent seismic Category II structures.

COL 3.7(5) The COL applicant is to perform seismic analysis for any site-specific seismic Category I dams, if necessary COL 3.7(6) The COL applicant is to perform seismic analysis of buried seismic Category I conduits and tunnels.

COL 3.7(7) The COL applicant is to perform seismic analysis for the seismic Category I above-ground tanks.

COL 3.7(8) The COL applicant that references the APR1400 design certification will determine whether essentially the same seismic response from a given earthquake is expected at each unit in a multi-unit site or each unit is to be provided with a separate set of seismic instruments.

COL 3.7(9) The COL applicant is to confirm details of the locations of the triaxial time-history accelerographs.

COL 3.7(10) The COL applicant is to identify the implementation milestones for the seismic instrumentation implementation program based on the discussion in Subsections 3.7.4.1 through 3.7.4.5.

COL 3.7(11) The COL applicant is to prepare a procedure for the post shutdown inspection and plant restart in accordance with the guidance of NRC RG 1.167.

COL 3.7B(1) The COL applicant is to perform the HRHF evaluation of safety-related equipment.

COL 3.7B(2) The COL applicant is to verify the applicability of evaluation of the items potential to HF sensitivity COL 3.7B(3) The COL applicant is to evaluate the HRHF response spectra for piping systems other than those within the scope of the graded approach.

1.8-10 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (7 of 39)

Item No. Description COL 3.8(1) The COL applicant is to perform concrete long-term material testing in a way which verifies physical properties of materials used during the design stage and the characteristics of long term deformation of concrete.

COL 3.8(2) The COL applicant is to provide the detailed design results and evaluation of the ultimate pressure capacity of penetrations, including the equipment hatch, personnel airlocks, electrical and piping penetrations in accordance with RG 1.216.

COL 3.8(3) The COL applicant is to provide detailed analysis and design procedure for the equipment hatch, personnel airlocks, and electrical penetrations.

COL 3.8(4) The COL applicant is to provide detailed analysis and design procedure for the transfer tube penetration assembly.

COL 3.8(5) The COL applicant is to provide the design of site-specific seismic Category I structures such as the essential service water building and the component cooling water heat exchanger building, essential service water conduits, component cooling water piping tunnel, and Class 1E electrical duct runs.

COL 3.8(6) The COL applicant is to evaluate any applicable site-specific loads such as explosive hazards in proximity to the site, projectiles and missiles generated from activities of nearby military installations, potential non-terrorism related aircraft crashes, and the effects of seiches, surges, waves, and tsunamis.

COL 3.8(7) The COL applicant is to perform the analysis and design of the steel plate for the new fuel storage pit.

COL 3.8(8) The COL applicant is to determine the environmental condition associated with the durability of concrete structures and provide the concrete mix design that prevents concrete degradation including the reactions of sulfate and other chemicals, corrosion of reinforcing bars, and influence of reactive aggregates.

COL 3.8(9) The COL applicant is to determine construction techniques to minimize the effects of thermal expansion and contraction due to hydration heat, which could result in cracking.

COL 3.8(10) For safety and serviceability of seismic Category I structures during the operation of the plant, the COL applicant is to provide appropriate testing and inservice inspection programs to examine the condition of normally inaccessible, below-grade concrete for signs of degradation and to conduct periodic site monitoring of ground water chemistry.

Inservice inspection of the accessible portion of concrete structures is also to be performed.

COL 3.8(11) The COL applicant is to verify that the coefficient of friction between the lean concrete and waterproofing membrane is greater than or equal to 0.55.

COL 3.8(12) The COL applicant is to provide reasonable assurance that the design criteria listed in Table 2.0-1 are met or exceeded.

COL 3.8(13) The COL applicant is to verify that the coefficient of friction between the lean concrete and the supporting medium at the site is equal to or higher than 0.55. In order to meet this requirement, the COL applicant is to determine the specific undulation pattern in Figure 3.8-27 for two perpendicular horizontal directions.

COL 3.8(14) The COL applicant is to confirm that uneven settlement due to construction sequence of the NI basemat falls within the values specified in Table 2.0-1.

1.8-11 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (8 of 39)

Item No. Description COL 3.8(15) The COL applicant is to provide a site-specific monitoring program and to monitor maximum vertical settlement, differential settlement, tilt, and angular distortion to ensure they are less than the criteria in Table 2.0-1, Table 3.8.5-12 thru Table 3.8.5-14, and section 3.8.5.4.2.2.d during construction and plant operation.

COL 3.8(16) The COL applicant is to provide testing and inservice inspection programs to examine inaccessible areas of concrete structures for degradation and to monitor groundwater chemistry.

COL 3.8(17) The COL applicant is to provide the following soil information for APR1400 site:

1) Elastic shear modulus and Poisson's ratio of the subsurface soil layers,

2) Consolidation properties including data from one-dimensional consolidation tests (initial void ratio, Cc, Ccr, OCR, and complete e-log p curves) and time-versus-consolidation plots,

3) Moisture content, Atterberg limits, grain size analyses, and soil classification,

4) Construction sequence and loading history, and

5) Excavation and dewatering programs.

COL 3.8(18) A detailed construction sequence analysis to determine the resulting construction settlements, including the various standard soils profiles (S01-S04, S06-S09) and sequencing of concrete pours for the NI common basemat (RCB and Auxiliary Building),

and superstructure model (Auxiliary Building, internal structures, and Shell & Dome), is presented in Section 3.8.5.4.2. A comparison of the four types of construction settlements (i.e. maximum vertical settlement, tilting settlement, maximum differential settlement between structures, and angular distortion) to the maximum criteria is summarized in Tables 3.8-12 through 3.8-14, and section 3.8.5.4.2.2.d.

The COL applicant should use the construction sequence settlement analysis given in Section 3.8.5.4.2, substituting site-specific soil layer conditions, to ensure that the four types of settlement criteria described in Table 3.8-12 thru Table 3.8-14, and section 3.8.5.4.2.2.d. are satisfied. An alternative construction sequence and settlement analysis may be performed by the COL applicant in response to 1) the inability to meet the settlement criteria described in Table 3.8-12 thru Table 3.8-14, and section 3.8.5.4.2.2.d.

using the DCD approach discussed in Section 3.8.5.4.2 or 2) Other site specific factors that may require a different construction plan and foundation sequence. However, in either case the COL applicant shall satisfy four types of settlement criteria described in Table 3.8-12 thru Table 3.8-14, and section 3.8.5.4.2.2.d..

1.8-12 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (9 of 39)

Item No. Description COL 3.8(19) The following items should be considered by the COL applicant:

1) The surveyed soil profiles will be developed.

2) Based on the surveyed soil characteristics, differences from the DCD soil profiles may exist. These differences may include:

a. Stiff or soft soil areas;

b. Different soil types (e.g, cohesive);

c. Potential for loss of cement in the mudmat;

d. Non-uniformity of soil layers, or

e. Other differences in the soil profile from the properties assumed in design certification. If any of these items and/or conditions are identified, then a site-specific evaluation1) shall be performed and checked for adequacy.

3) The time (i.e, short term and long term) instantaneous settlement and time-consolidation effects shall be evaluated in accordance with surveyed soil profiles regardless if a site-specific evaluation is needed under Item 2) above. The bearing pressure shall be checked to demonstrate acceptability with the acceptance criteria in DCD Table 2.0-1. Settlements shall be checked in Table 3.8-12 thru Table 3.8-14, and section 3.8.5.4.2.2.d.

4) The COL applicant will build the seismic Category I structure according to the construction sequence used in the site-specific construction sequence analysis.

5) If a site-specific evaluation1) is required, the COL applicant should perform a construction sequence analysis based on the site-specific parameters. If the settlement including results of construction sequence analysis exceeds the acceptance criteria described in Table 3.8-12 thru Table 3.8-14, and section 3.8.5.4.2.2.d., the construction sequence will be modified to meet the acceptance criteria described in Table 3.8-12 thru Table 3.8-14, and section 3.8.5.4.2.2.d. by COL applicant.

6) The effect on the design of seismic Category I structures due to construction sequence analysis shall be accounted for by the COL applicant.

COL 3.8(20) The COL applicant shall perform site-specific evaluations if the shear wave velocity is less than the shear wave velocity profile used in the various basemat evaluations for design certification. The site-specific evaluations ([settlement (maximum vertical displacement, tilt, differential settlement between structures, angular distortion), overturning, and sliding])

and 3D FEM global analysis for basemat design of seismic Category I structures shall be performed using the site-specific parameters (measured Estatic, Edynamic consistent with soil strain assumed in SSI analysis) and the methodology described in Subsection 3.8.5 and Technical report APR1400-E-S-NR-14006-P, Subsection 4.

COL 3.8(21) The COL applicant is to confirm that the parking position of the crane and trolley when the crane is not being used is: location of polar crane: Az.280º, trolley location: 12ft 7in away from end of east part. The COL applicant is to confirm that this requirement is included in the technical specification of the COL application for the use of the polar crane.

1) evaluation includes basemat and superstructure design (forces/stresses), settlement evaluations, soil bearing pressure evaluation, and stability evaluation.

1.8-13 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (10 of 39)

Item No. Description COL 3.9(1) The COL applicant is to provide the inspection results for the APR1400 reactor internals classified as non-prototype Category I in accordance with RG 1.20.

COL 3.9(2) The COL applicant is to identify the site-specific active pumps.

COL 3.9(3) The COL applicant is to provide a full description of the IST program (including PST and MOV testing) for pumps, valves and dynamic restraints that will be administratively controlled such that the applicable requirements of the ASME OM Code edition and addenda are incorporated in the IST program.

COL 3.9(4) The COL applicant is to provide an IST program including the type of testing and frequency of site-specific pumps subject to IST in accordance with the ASME OM Code and Table 3.9-13 COL 3.9(5) The COL applicant is to provide an IST program including the type of testing and frequency of any site-specific valves subject to IST in accordance with the ASME OM Code and Table 3.9-13 COL 3.9(6) The COL applicant is to provide a table listing all safety-related components that use snubbers in their support systems.

COL 3.9(7) The COL applicant is to review the detailed analysis of mechanical tanks, including the effects of fluid sloshing.

COL 3.10(1) The COL applicant is to provide documentation that the designs of seismic Category I SSCs are analyzed for OBE, if OBE is higher than 1/3 SSE.

COL 3.10(2) The COL applicant is to investigate if site-specific spectra generated for the COLA exceed the APR1400 design spectra in the high-frequency range. Accordingly, the COL applicant is to provide reasonable assurance of the functional performance of vibration-sensitive components in the high-frequency range.

COL 3.10(3) The COL applicant is to develop the equipment seismic qualification files that summarize the components qualification, including a list of equipment classified as seismic Category I in Table 3.2-1 and seismic qualification summary data sheets (SQSDS) for each piece of seismic Category I equipment.

COL 3.10(4) The COL applicant is to perform equipment seismic qualification for seismic Category I equipment and provide milestones and completion dates of equipment seismic qualification program.

COL 3.11(1) The COL applicant is to identify and qualify the site-specific mechanical, electrical, I&C, and accident monitoring equipment specified in RG 1.97.

COL 3.11(2) The COL applicant is to identify the nonmetallic parts of mechanical equipment in procurement process.

COL 3.11(3) The COL applicant is to operational address aspects for maintaining the environmental qualification status of components after initial qualification.

1.8-14 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (11 of 39)

Item No. Description COL 3.11(4) The COL applicant is to provide a full description of the environmental qualification of mechanical and electrical equipment program.

COL 3.11(5) The COL applicant is to document the qualification test results and qualification status in an auditable file for each type of equipment in accordance with the requirements 10 CFR 50.49(j).

COL 3.11(6) The COL applicant is to describe the EQP implementation milestones based on the APR1400 EQP.

COL 3.11(7) The COL applicant is to provide the room number designation for equipment with unidentified rooms in Table 3.11-2, and ensure that the equipment is located within the environmental and radiation conditions specified.

COL 3.12(1) If COL applicant finds it necessary to route ASME Class 1, 2, or 3 piping systems outside the structure, the wind and/or tornado load must be included in the plant design basis loads considering the site-specific loads.

COL 3.12(2) The COL applicant will implement the monitoring program during the first preoperational testing and continue to monitor by using the fatigue monitoring system during the first cycle of operation to verify the design transients used in the structural design of the surge line.

COL 3.13(1) The COL applicant is to maintain quality assurance records including CMTRs on ASME Section III Class 1, 2, and 3 component threaded fasteners in accordance with the requirements of 10 CFR 50.71.

COL 3.13(2) The COL applicant is to submit the preservice and inservice inspection programs for ASME Section III Class 1, 2, and 3 component threaded fasteners to the NRC prior to performing the inspections.

COL 5.2(1) The COL applicant is to address the addition of ASME Code cases that are approved in NRC RG 1.84 at the time of the application.

COL 5.2(2) The COL applicant is to address the ASME Code cases approved in NRC RG 1.147 at the time of the application and invoked for the ISI program of specific plant.

COL 5.2(3) The COL applicant is to address the ASME Code cases approved in NRC RG 1.192 at the time of the application and invoked for operation and maintenance activities of a specific plant.

COL 5.2(4) The COL applicant is to address the material specifications, which are not shown in Table 5.2-2, as necessary.

COL 5.2(5) The COL applicant will review and confirm at the time of COL submittal based on the latest revision of EPRI PWR Primary Side Water Chemistry Guidelines.

COL 5.2(6) The COL applicant is to address the actual, as-procured, fracture toughness data of the RCPB materials to the staff at a predetermined time by an appropriate method.

COL 5.2(7) The COL applicant is to submit the actual, as-procured yield strength of the austenitic stainless steel materials used in RCPB to the staff at a predetermined time agreed-upon by the regulatory body.

COL 5.2(8) The COL applicant is to provide and develop the implementation milestones for the inservice inspection and testing program for the RCPB, in accordance with ASME Code Section XI and 10 CFR 50.55a.

1.8-15 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (12 of 39)

Item No. Description COL 5.2(9) The COL applicant is to address the provisions to accessibility of Class 1 components for ISI if the design of the APR1400 Class 1 component is changed from the DCD design.

COL 5.2(10) The COL applicant is to provide the list of Code exemptions in the ISI program of the specific plants, if it exists.

COL 5.2(11) The COL applicant is to prepare and implement a boric acid corrosion (BAC) prevention program compliant with Generic Letter 88-05.

COL 5.2(12) The COL applicant is to prepare the preservice inspection and testing program.

COL 5.2(13) The COL applicant is to address and develop milestones for preparation and implementation of the procedure for operator responses to prolonged low level leakage.

COL 5.2(14) The COL applicant is to address the portions of the later editions and addenda of ASME Code.

COL 5.3(1) The COL applicant is to provide a reactor vessel material surveillance program for a specific plant.

COL 5.3(2) The COL applicant is to provide plant-specific surveillance capsule data to benchmark APR1400-Z-A-NR-14015-P and demonstrate applicability to the specific plant.

COL 5.3(3) The COL applicant is to develop P-T limit curves based on plant-specific data.

COL 5.3(4) The COL applicant is to verify the RTPTS value and the USE at EOL based on plant-specific material property and neutron fluences.

COL 5.3(5) The COL applicant is to provide and develop the inservice inspection and testing program for the RCPB, in accordance with ASME Section XI and 10 CFR 50.55a.

COL 5.4(1) The COL applicant is to prepare operational procedures and maintenance programs as related to leak detection and contamination control of RCS.

COL 5.4(2) The COL applicant is to maintain complete documentation of system design, construction, design modifications, field changes, and operations. Documentation requirements are included as a COL information item.

COL 5.4(3) The COL applicant is to prepare operational procedures and maintenance programs as related to leak detection and contamination control of SCS.

COL 5.4(4) The COL applicant is to maintain complete documentation of system design, construction, design modifications, field changes, and operations of SCS.

COL 5.4(5) The COL applicant is to verify the as-built RV support material properties and 60-year neutron fluence.

COL 5.4(6) The COL applicant is to provide a procedure to protect the plant personnel from harsh environment and lengthen the work time within containment during hatch closure.

COL 5.4(7) The COL applicant is to prepare an inspection and monitoring program for the cladding material integrity of SG channel heads.

COL 6.1(1) The COL applicant is to managing nonconforming coatings that do not meet the service Level I requirements in accordance with ASTM D7491-08.

COL 6.1(2) The COL applicant is to describe the coating program and to identify the implementation milestones for the coating program.

1.8-16 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (13 of 39)

Item No. Description COL 6.1(3) The COL applicant is to identify the amount of cable insulation/jacket inside containment as indicated in Table 6.1-4.

COL 6.2(1) The COL applicant is to provide the weld filler material in the supplier specification.

COL 6.3(1) The COL applicant is to prepare operational procedures and maintenance programs as related to leak detection and contamination control.

COL 6.3(2) The COL applicant is to maintain complete documentation of system design, construction, design modifications, field changes, and operations.

COL 6.4(1) The COL applicant is to provide automatic and manual operating procedures for the control room HVAC system, which are required in the event of a postulated toxic gas release.

COL 6.4(2) The COL applicant is to provide the details of specific toxic chemicals of mobile and stationary sources and evaluate the MCR habitability based on the recommendations in NRC RG 1.78 to meet the requirements of TMI Action Plan Item III.D.3.4 and GDC 19.

COL 6.4(3) The COL applicant is to identify and develop toxic gas detection requirements to protect the operators and provide reasonable assurance of the MCR habitability. The number, locations, sensitivity, range, type, and design of the toxic gas detectors are to be developed by the COL applicant.

COL 6.5(1) The COL applicant is to provide the operational procedures and maintenance program as related to leak detection and contamination control.

COL 6.5(2) The COL applicant is to maintain the complete documentation of system design, construction, design modifications, field changes, and operations.

COL 6.6(1) The COL applicant is to identify the implementation milestones for ASME Section Xl inservice inspection program for ASME Code Section III Class 2 and 3 components.

COL 6.6(2) The COL applicant is to provide the preservice inspection program and the inservice inspection program.

COL 6.6(3) The COL applicant is to address the provisions to accessibility of Class 2 or 3 components for ISI if the design of the APR1400 Class 2 or 3 components is changed from the DCD Design.

COL 6.6(4) The COL applicant is to identify the implementation milestone for the augmented inservice inspection program and provide a description of the augmented in-service inspection which is to be conducted..

COL 6.8(1) The COL applicant is to provide the operational procedures and maintenance program for leak detection and contamination control.

COL 6.8(2) The COL applicant is to provide the preparation of cleanliness, housekeeping, and foreign materials exclusion program.

COL 6.8(3) The COL applicant is to maintain the complete documentation of system design, construction, design modifications, field changes, and operations.

COL 6.8(4) The COL applicant is to confirm that the RMI is one of the tested RMIs in NUREG/CR-6808.

1.8-17 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (14 of 39)

Item No. Description COL 6.8(5) The COL applicant is to confirm that the IRWST sump strainer has the total strainer head loss less than the allowable head loss (0.61 m (2ft)).

COL 6.8(6) The COL applicant is responsible for the establishment and implementation of the Maintenance Rule program in accordance with 10 CFR 50.65.

COL 7.1(1) The COL applicant is to provide the software installation plan, the software training plan, and the software operation and maintenance plan for the safety I&C systems, as described in the Software Program Manual Technical Report.

COL 7.1(2) The COL applicant is to provide justifiable software reliability data for software used in the digital I&C systems (i.e., PPS and DPS).

COL 7.2(1) The COL applicant is to provide site-specific CPCS startup test requirement COL 7.5(1) The COL applicant is to provide a description of the site-specific AMI variables such as wind speed, and atmosphere stability temperature difference.

COL 7.5(2) The COL applicant is to provide a description of the site-specific EOF.

COL 7.5(3) The COL applicant is to provide the bases document accounting for measurement uncertainties for the EOP action points.

COL 8.2(1) The COL applicant is to identify the circuits from the transmission network to the onsite electrical distribution system that are supplied by two physically independent circuits.

COL 8.2(2) The COL applicant is to provide information on the location of rights-of-way, transmission towers, voltage level, and length of each transmission line from the site to the first major substation that connects the line to the transmission network.

COL 8.2(3) The COL applicant is to describe the switchyard voltage related to the transmission system provider/operator (TSP/TSO) and the formal agreement between the nuclear power plant and the TSP/TSO. The COL applicant is to describe the capability and the analysis tool of the TSP. The COL applicant is also to describe the protocols for the plant to remain cognizant of grid vulnerabilities.

COL 8.2(4) The COL applicant is to describe and provide layout drawings of the circuits connecting the onsite distribution system to the preferred power supply.

COL 8.2(5) The COL applicant is to describe the site-specific design for the switchyard equipment, including breaker arrangement, electrical schematics of breaker control system, protective devices and their settings, and auxiliary power supplies (ac and dc) for control and protection.

COL 8.2(6) The COL applicant is to provide a high-impedance ground fault detection feature that provides an alarm in the MCR and RSR upon detection of a high-impedance ground fault at the primary side of MT or SATs.

COL 8.2(7) The COL applicant is to provide an FMEA of the switchyard components to assess the possibility of simultaneous failure of both circuits as a result of single events. In addition, the COL applicant is to provide the results of grid stability analyses to demonstrate that the offsite power system does not degrade the normal and alternate preferred power sources to a level where the preferred power sources do not have the capacity or capability to support the onsite Class 1E electrical distribution system in performing its intended safety function.

1.8-18 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (15 of 39)

Item No. Description COL 8.2(8) The COL applicant is to determine the specific type of the OPDP system which properly address and meet the requirements of B.1. and B.2. of Branch Technical Position (BTP) 8-9, taking into account the site specific design configuration, installation condition, (field) performance testing and qualification status, and operation experiences of the OPDP system.

The COL applicant is also to provide the detailed design of the OPDP system selected for the APR1400 site.

The COL applicant is to perform a field simulation on the site specific design of the offsite power system to ensure that the settings of the OPDP system are adequate and appropriate for the site.

COL 8.2(9) The COL applicant is to describe how testing is performed on the offsite power system components and identify the potential effects that must be considered during testing.

COL 8.2(10) The COL applicant is to provide the required number of immediate access circuits from the transmission network.

COL 8.3(1) The COL applicant is to provide and to design a mobile GTG and its support equipment.

COL 8.3(2) The COL applicant is to provide a bus transfer study of the onsite power system. Based on the bus transfer study, the COL applicant is also to provide final relay selection and settings for the bus transfer.

COL 8.3(3) The COL applicant is to establish procedures to monitor and maintain EDG reliability during plant operations to verify the selected reliability level target is being achieved as intended in RG 1.155.

COL 8.3(4) The COL applicant is to describe and provide detailed ground grid and lightning protection.

COL 8.3(5) The COL applicant is to conduct periodic inspection and testing of the protection devices for the EPA conductors. All circuit breakers for the EPA conductors shall be inspected and tested in 60 months, low voltage circuit breaker overcurrent protection devices for the EPA conductors shall be inspected and tested once per 18 months for 10 % of each type of circuit breakers, and overcurrent relay for medium voltage circuit breakers for the EPA conductors shall be inspected and tested once per 18 months.

COL 8.3(6) The COL applicant is to provide testing, inspection, and monitoring programs for detecting insulation degradation of underground and inaccessible power cables within the scope of 10 CFR 50.65.

COL 8.3(7) The COL applicant is to establish Administrative Program(s), including application of dedicated cable and raceway management database tool as necessary, which is (are) developed on the basis of the cable and raceway numbering system to efficiently manage cable routing and cable termination and verify that the cable design fulfills the acceptance criteria (i.e., separation, filling criteria, and ampacity).

COL 8.3(8) The COL applicant is to provide the detailed design of the cathodic protection system as applicable to the site conditions.

COL 8.3(9) The COL applicant is to provide protective device coordination.

COL 8.3(10) The COL applicant is to provide the analysis and underlying assumptions used to demonstrate adequacy for insulation coordination of surge and lightning protection.

1.8-19 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (16 of 39)

Item No. Description COL 8.3(11) The COL applicant is to provide the analysis and underlying assumptions used to demonstrate adequacy for power quality limits (harmonic distortion).

COL 8.3(12) The COL applicant is to provide the analysis for the station and switchyard grounding system with underlying assumptions, based on the site-specific parameters including soil resistivity and site layout.

COL 8.3(13) The COL applicant is to develop the maintenance program to optimize the life and performance of the batteries.

COL 8.3(14) The COL applicant is to provide short circuit analysis of onsite dc power system with actual data.

COL 8.3(15) The COL applicant is to describe any special features of the design that would permit online replacement of an individual cell, group of cells, or entire battery.

COL 8.4(1) The COL applicant is to validate the SBO coping duration according to the method specified in RG 1.155.

COL 8.4(2) The COL applicant is to identify local power sources and transmission paths that could be made available to resupply power to the plant following the loss of a grid or the SBO.

COL 8.4(3) The COL applicant is to specify the specific parameters for monitoring, alarms, mechanical and electrical trip for testing, and emergency trips.

COL 8.4(4) The COL applicant is to develop detailed procedures for manually aligning the alternate AC power supply when two (Trains A and B) of the four diesel generators are unavailable during a loss of offsite power event.

COL 9.1(1) The COL applicant shall create a weld mockup that realistically represents the spent fuel storage racks in case of the neutron absorbing material to be installed in the sheath cavity prior to sheath welding. During welding of the mockup, the applicant shall monitor the temperature of the fuel rack to determine the maximum temperature which the neutron absorbing material is exposed to. The WPS used for the mockup shall be used for fabrication and the fabrication welds shall not exceed the heat input of the mockup welds.

If the heat input is exceeded during fabrication, the neutron absorbing material shall be treated as non-conforming. The non-conformance would be acceptable following requalification which includes a mockup to determine if the new heat input results in neutron absorbing material being exposed to temperatures exceeding 900 °F.

COL 9.1(2) The COL applicant is to prepare a periodic inspection program of spent fuel rack integrity to ensure that the design basis material and geometric assumptions remain valid during the operating life of the plant.

COL 9.1(3) The COL applicant is to perform a confirmatory structural dynamic and stress analysis for the new and spent fuel racks including reconciliation of loads imposed by the new and spent fuel racks on the new fuel pit and the spent fuel pool structure, considering the site specific condition such as a safe shutdown earthquake.

COL 9.1(4) The COL applicant is to provide plant procedures and administrative control for fuel handling activities over the spent fuel pool.

COL 9.1(5) The COL applicant is to provide post seismic event inspection procedures to measure gaps between the new and spent fuel storage racks.

1.8-20 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (17 of 39)

Item No. Description COL 9.1(6) The COL applicant is to provide operational procedures and maintenance programs for the inspection, calibration, testing, and maintenance of the SFP leak detection provision (sight glasses), pool water temperature, pool water level, SFP area radiation monitor, and the SFP steel liner. The COL applicant is also to provide the inspection interval for the maintenance program. In addition, the COL applicant is to provide operational procedures and maintenance programs for the design features implemented for SFP contamination control, including heat exchanger seals, epoxy coating, and SFP filters and demineralizers. The contamination control procedures and programs can be integrated into an overall plant-wide RG 4.21 program following the guidance from NEI 08-08A.

COL 9.1(7) The COL applicant is to maintain complete documentation of system design and system design modifications.

COL 9.1(8) The COL applicant is to provide plant procedures for preventing and mitigating inadvertent refueling pool drain down events, maintenance procedures for the maintenance and inspection of refueling pool seal, and emergency response procedures for the proper measures during pool drain down events.

COL 9.1(9) The COL applicant is to provide plant operating procedure guidelines for preoperational load testing and checks of interlocks, blocks, hoisting cables, control circuity and lubrication of fuel handling equipment.

COL 9.1(10) The COL applicant is to address the load-handling procedures. Load-handling procedures are established for component handling procedures and plant operating procedures in accordance with ASME B30.2. ASME B30.2 requires establishing component handling procedures that include (1) a safe load path for lifting heavy loads to perform special handling component inspections, (2) acceptance criteria prior to lift, and (3) use of steps and proper sequence in handling the load. ASME B30.2 requires plant operating procedure guidelines that include appropriate crane operator training and crane inspections. ASME B30.2 also requires that the load-handling procedures include preparing operating procedures for preoperational load testing and checkouts of interlocks, brakes, hoisting cables, control circuitry, and lubrication of OHLHS equipment.

COL 9.2(1) The COL applicant is to provide the UHS-related systems such as blowdown, chemical injection, and makeup water system.

COL 9.2(2) The COL applicant is to develop procedures for system filling, venting, and operational procedures to minimize the potential for water hammer; to analyze the system for water hammer impacts; to design the piping system to withstand potential water hammer forces; and to analyze inadvertent water hammer events in accordance with NUREG-0927 in the ESWS.

COL 9.2(3) The COL applicant is to develop layout of the site-specific portion of the system to minimize the potential for water hammer in the ESWS.

COL 9.2(4) The COL applicant is to (1) to determine required pump design head, using pressure drop from the certified design portion of the plant and adding site-specific head requirements, (2) determine pump shutoff head to establish system design pressure, which is not to exceed APR1400 system design pressure, and (3) evaluate potential for vortex formation at the pump suction based on the most limiting applicable conditions in the ESWS.

1.8-21 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (18 of 39)

Item No. Description COL 9.2(5) The COL applicant is to provide the detailed information for the ESWS piping material including the basis for determining that the materials being used are appropriate for a given site location and for the fluid properties that apply.

COL 9.2(6) The COL applicant is to determine the design details of the backwashing line, vent line, and their discharge locations in the ESWS.

COL 9.2(7) The COL applicant is to provide operational procedures and maintenance programs as related to leak detection and contamination control in the ESWS.

COL 9.2(8) The COL applicant is to provide the evaluation of the ESW pump at the high and low water levels of the UHS. In the event of approaching low UHS water level, the COL applicant is to develop a recovery procedure.

COL 9.2(9) The COL applicant is to provide measures to prevent long-term corrosion and organic fouling that may degrade system performance in the ESWS.

COL 9.2(10) The COL applicant is to evaluate the need and design and install freeze protection in the ESWS if required.

COL 9.2(11) The COL applicant is to conduct periodic inspection, monitoring, maintenance, performance, and functional testing of the ESWS and UHS piping and components, including the heat transfer capability of the CCW heat exchangers based on GL 89-13 and GL 89-13 Supplement 1.

COL 9.2(12) The COL applicant is to develop procedures for water systems filling, venting, keeping the system full, and operation to minimize the potential for water hammer in accordance with NUREG-0927 in the CCWS.

COL 9.2(13) The COL applicant is to evaluate the effects of exposure to subfreezing environments for the CCWS piping and SSCs located outside of plant buildings, and, if required, to provide measures to protect from the subfreezing effects.

COL 9.2(14) The COL applicant is to provide operational procedures and maintenance programs as related to leak detection and contamination control in the CCWS.

COL 9.2(15) The COL applicant is to maintain complete documentation of system design, construction, design modifications, field changes, and operations in the CCWS.

COL 9.2(16) The COL applicant is to include a site-wide radiological environmental monitoring program to monitor environmental contamination in the CCWS.

COL 9.2(17) The COL applicant is to determine all state and local departments of health and environmental protection standards to be applied and followed for the domestic water system.

COL 9.2(18) The COL applicant is to determine the source of domestic water to the site and the necessary required treatment plant.

COL 9.2(19) The COL applicant is to confirm the sizing of domestic water tanks and associated pumps, if used.

COL 9.2(20) The COL applicant is to confirm whether the sanitary waste is sent to an onsite treatment facility or the city sewage system.

1.8-22 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (19 of 39)

Item No. Description COL 9.2(21) The COL applicant is to provide the UHS-related design information based on the site characteristics, including meteorological conditions.

COL 9.2(22) The COL applicant is to provide the location and design of the ESW building, and makeup water source.

COL 9.2(23) The COL applicant is to provide isolation between the UHS and the non-safety-related systems.

COL 9.2(24) The COL applicant is to provide the design of UHS cooling tower basin so the minimum water level will provide adequate NPSH to ESW pumps under accident conditions.

COL 9.2(25) The COL applicant is to provide the non-safety-related makeup water source and capacity for normal operation loss and evaporation in the UHS.

COL 9.2(26) The COL applicant is to specify the following UHS chemistry requirements for bio-fouling and chemistry control:

a. A chemical injection system to provide non-corrosive, non-scale-forming conditions to limit biological film formation

b. The type of biocide, algaecide, pH adjuster, corrosion inhibitor, scale inhibitor, and silt dispersant, if necessary to maintain system performance based on site conditions.

COL 9.2(27) The COL applicant is to verify the piping layout of the ESWS and UHS to prevent water hammer and develop operating procedures to provide reasonable assurance that the ESWS and UHS water pressure are above saturation conditions for all operating modes.

COL 9.2(28) The COL applicant is to develop maintenance and testing procedures to monitor debris buildup and flush out and to remove the debris in the UHS.

COL 9.2(29) The COL applicant is to evaluate the potential wind and recirculation effects of cooling towers based on meteorological condition.

COL 9.2(30) The COL applicant is to provide the material specifications for piping, valves, and fittings of the UHS system based on site-specific conditions and meteorological conditions.

COL 9.2(31) The COL applicant is to provide the evaluation of maximum evaporation and other losses based on the site-specific conditions and meteorological conditions in the UHS.

COL 9.2(32) The COL applicant is to provide the detailed evaluation for UHS capability with consideration of site-specific conditions and meteorological data in the UHS.

COL 9.2(33) The COL applicant is to provide chemical and blowdown to prevent biofouling and long-term corrosion, considering site water quality in the UHS.

COL 9.2(34) The COL applicant is to provide the inspection and testing of the UHS to demonstrate that fouling and degradation mechanisms applicable to the site are effectively managed to maintain acceptable heat sink performance and integrity.

COL 9.2(35) The COL applicant is to provide the design provisions to permit the preoperational and in-service testing and inspection based on the type of UHS.

COL 9.2(36) The COL applicant is to provide the alarms, instrumentation, and controls required for the safety-related functions of the UHS.

1.8-23 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (20 of 39)

Item No. Description COL 9.2(37) The COL applicant is to develop the following procedures for the water system: filling, venting, keeping it full, and operating it to minimize the potential for water hammer. The COL applicant is also to analyze the system for water hammer impacts, design the piping system to withstand potential water hammer forces, and analyze inadvertent water hammer events in the ECWS in accordance with NUREG-0927.

COL 9.2(38) The COL applicant is to confirm that there are no departures and shall meet the interface requirements (i.e., cooling duties and temperature requirements, piping and control interface)

COL 9.2(39) The COL applicant is to provide operational procedures and maintenance programs as related to leak detection and contamination control in accordance with RG 4.21.

COL 9.2(40) The COL applicant is to include a site-wide radiological environmental monitoring program to monitor both the horizontal and vertical variability of the onsite hydrogeology and the potential effects of the construction and operation of the plant.

COL 9.2(41) The COL applicant is to maintain complete documentation of system design and any site specific design modifications during the COL application, for the features for contamination control, in accordance with RG 4.21, Subsection A-3 to facilitate decommissioning.

COL 9.3(1) The COL applicant is to provide the supply systems of the nitrogen gas subsystem, the hydrogen subsystem, the carbon dioxide subsystem, and the breathing air systems.

COL 9.3(2) The COL applicant is to provide operational procedures and maintenance programs as related to leak detection and contamination control.

COL 9.3(3) The COL applicant is to maintain complete documentation of system design, construction, design modifications, field changes, and operations.

COL 9.3(4) The COL applicant is to provide the flow diagram of turbine generator building drain system and the interconnection from the auxiliary boiler building sump, and the flow diagram of CCW heat exchanger building drain system to LWMS or turbine generator building (TGB) sump.

COL 9.3(5) The COL applicant is to provide connection provisions at the nearest accessible area to the valve and sight glass room of the IRWST leakage pipe line for detecting and cleaning blockage due to crystallized boron inside the leakage collection channel and pipes.

COL 9.3(6) The COL applicant is to prepare the site radiological environmental monitoring program.

COL 9.3(7) The COL applicant is to provide primary side water chemistry threshold values and recommended operator actions for chemistry excursions in compliance with the latest version of the EPRI PWR Primary Water Chemistry Guidelines in effect at the time of COLA submittal. The COL applicant is to establish the operational water chemistry program six months before fuel load.

COL 9.3(8) The sump for the outdoor tanks (HUT and BAST) is located within the structural walls surrounding the tanks. The sump is enclosed with a sump cover, similar to the indoor sumps, which serve to protect the sump from the elements. The outdoor sump design is site-specific. The COL applicant is to determine the need and the provision of freeze protection for the outdoor sumps.

COL 9.4(1) The COL applicant is to provide the capacities of heating coils in the safety-related air handling units and cooling and heating coils in the non-safety-related air handling units affected by site-specific conditions.

COL 9.4(2) The COL applicant is to provide the interval of reopening the closed outside air intake isolation dampers by considering the durability of the isolation dampers and the site-specific meteorological data and to consider the radiological impacts if a higher frequency is determined.

1.8-24 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (21 of 39)

Item No. Description COL 9.4(3) The COL applicant is to provide the system design information of ESW building and CCW heat exchanger building HVAC system including flow diagram, if the ESW building and CCW heat exchanger building require the HVAC system.

COL 9.4(4) The COL applicant is to provide the capacities of heating coils of electric duct heaters affected by site-specific conditions.

COL 9.4(5) The COL applicant is to establish operational procedures and maintenance programs as related to leak detection and contamination control.

COL 9.5(1) The COL applicant is to establish a fire protection program, including organization, training, and qualification of personnel, administrative controls of combustibles and ignition sources, firefighting procedures, and quality assurance.

COL 9.5(2) The COL applicant is to address the design and fire protection aspects of the facilities, buildings and equipment, and a fire protection water supply system, which are site specific and/or are not a standard feature of the APR1400.

COL 9.5(3) The COL applicant is to describe the provided apparatus for plant personnel and fire brigades such as portable fire extinguishers, self-contained breathing apparatus, and radio communication systems.

COL 9.5(4) The COL applicant is to address the final FHA and FSSA based on the final plant design, including a detailed post-fire safe-shutdown circuit analysis.

COL 9.5(5) The COL applicant is to provide a description of the offsite communication system that interfaces with the onsite communication system, including type of connectivity, radio frequency, normal and backup power supplies, and plant security system interface.

COL 9.5(6) The COL applicant is to provide the LAN and VPN system.

COL 9.5(7) The COL applicant is to provide the fire brigade radio systems.

COL 9.5(8) The COL applicant is to provide the local law enforcement communications including dedicated conventional telephone and radio-transmitted two-way communication system.

COL 9.5(9) The COL applicant is to provide the emergency offsite communication system including dedication hotline, local law enforcement radio equipment, and wireless communication system.

COL 9.5(10) The COL applicant is to provide the security radio system that consists of a base unit, mobile units, and portable units.

COL 9.5(11) The COL applicant is to provide security lighting for the protected area barrier and isolation zones and normal offsite power for all the exterior security lighting systems.

COL 9.5(12) The COL applicant is to identify the applicable codes and standards for the external coating on the underground piping in accordance with NRC RG 1.137.

COL 9.5(13) The COL applicant is to specify that adequate and acceptable sources of fuel oil are available, including the means of transporting and recharging the fuel storage tank, following a design basis accident.

COL 9.5(14) The COL applicant is to provide a reliable starting method for the AAC GTG.

1.8-25 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (22 of 39)

Item No. Description COL 9.5(15) The COL applicant is to provide the system design information of AAC GTG building HVAC system including flow diagram, if the AAC GTG building requires the HVAC system.

COL 10.2(1) The COL applicant is to identify the turbine vendor and model.

COL 10.2(2) The COL applicant is to identify how the functional requirements as described in Subsection 10.2.2.3.2 for the overspeed protection system are met and provide a schematic(s) of the TGCS and overspeed protection systems that show the entire system end-to-end and all discrete components and interfaces (e.g., sensors, power supplies, control devices, manual emergency trips, the device that eventually drains the hydraulic/air fluid from turbine control valves). The schematics and descriptive information provided once a turbine design is selected shall be sufficient to allow assessment of the TGCS and overspeed systems' ability to withstand a single failure without loss of function (i.e., redundancy), resistance to common cause failure (i.e., diversity as provided by electrical and mechanical overspeed trips), and resistance of propagation of a failure to another trip channel (i.e., independence)

COL 10.2(3) The COL applicant is to provide a description of how the turbine missile probability analysis conforms with Subsection 10.2.3.6 to ensure that requirements for protection against turbine missiles (e.g., applicable material properties, method of calculating the fracture toughness properties per SRP Section 10.2.3 Acceptance Criteria, preservice inspections, in-service tests and inspections) will be met. If the turbine vendor has performed a turbine missile analysis that has been reviewed and approved by the NRC for a rotor design relevant to the COL applicants selected design, then the COL should reference the analysis. If an approved analysis is not available, then the COL applicant shall prepare and reference an analysis that provides confidence that the final analysis performed with as-built properties, when available, will be sufficient to demonstrate assurance of turbine rotor integrity.

COL 10.2(4) The COL applicant shall specify the turbine rotor material properties for chosen turbine vendor and applicable for the specific rotor designs. The COL applicant shall specify the turbine rotor material properties (in terms of the 50% FATT and Charpy V-notch energy) for the chosen turbine vendor and applicable for the specific rotor designs. Any deviation from material properties in SRP 10.2.3, revision in effect on date of regulatory applicability for COL application, shall be identified and justified.

COL 10.2(5) The COL applicant shall provide the site-specific turbine rotor inservice inspection program and inspection interval, including the turbine valve test and inspection program and test and inspection frequency consistent with the manufacturer's turbine missile analysis.

COL 10.3(1) The COL applicant is to provide operating and maintenance procedures in accordance with NUREG-0927 and a milestone schedule for implementation of the procedure.

COL 10.3(2) The COL applicant is to establish operational procedures and maintenance programs as related to leak detection and contamination control.

COL 10.3(3) The COL applicant is to provide secondary side water chemistry threshold values and recommended operator actions for chemistry excursions in compliance with the latest version of the EPRI PWR Secondary Water Chemistry Guidelines in effect at the time of COLA submittal. The COL applicant is to establish the operational water chemistry program six months before fuel load.

1.8-26 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (23 of 39)

Item No. Description COL 10.3(4) The COL applicant is to provide the description about the material specifications for components between 1) the high pressure turbine and the moisture separator reheater and 2) the moisture separator reheater and the low pressure turbine when the T/G design is selected.

The COL applicant is also to specify that the pipe thickness is adequate for the plant design life in terms of FAC in place of the components between 1) the high pressure turbine and the moisture separator reheater and 2) the moisture separator reheater and the low pressure turbine when the T/G design is selected.

COL 10.3(5) The COL applicant is to provide a description of the FAC monitoring program. The description is to address consistency with GL 89-08 and NSAC-202L-R3 and provide a milestone schedule for implementation of the program. The program shall incorporate the conditions of 10 CFR 50.55a(b)(5) on ASME Code Case N-597-2.

COL 10.3(6) The COL applicant is to provide material specifications that will be utilized for ASME Section III components.

COL 10.4(1) The COL applicant is to establish operational procedures and maintenance programs for leak detection and contamination control COL 10.4(2) The COL applicant is to maintain the complete documentation of system design, construction, design modifications, field changes, and operations COL 10.4(3) The COL applicant is to provide the location and design of the cooling tower, basin, and CW pump house, if used.

COL 10.4(4) The COL applicant is to confirm that the water hammer events are bounded by the system design pressure value with a hydraulic transient analysis or otherwise demonstrate that the design is acceptable to satisfy GDC 4 in regard to the design provisions that are implemented to accommodate the effects of discharging water that could result from a malfunction or failure of a component or piping in the system.

COL 10.4(5) The COL applicant is to provide elevation drawings COL 10.4(6) The COL applicant is to address the design features for the prevention of contamination COL 10.4(7) The COL applicant is responsible for provisions of temporary shielding, if required, and mobile equipment, including spent resin fill-head for packaging of the contaminated spent resin, provisions of temporary storage, and shipment of packaged contaminated CPS spent resin for off-site treatment and/or disposal.

COL 10.4(8) The COL applicant is to provide operating and maintenance procedures in accordance with NUREG-0927 and a milestone schedule for implementation of the procedures.

COL 10.4(9) The COL applicant is to describe the nitrogen or equivalent system design for SG drain mode.

COL 10.4(10) The COL applicant is to prepare the Site Radiological Environmental Monitoring Program COL 10.4(11) The COL applicant is to develop procedures to perform periodic testing or maintenance, including independent verification in accordance with NUREG-0635.

COL 10.4(12) The COL applicant is to determine the wet bulb temperature correction factor to account for potential interference and recirculation effects 1.8-27 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (24 of 39)

Item No. Description COL 11.2(1) The COL applicant is to prepare the site-specific ODCM in accordance with NEI 07-09A.

COL 11.2(2) The COL applicant is to prepare operational procedures and programs related to operations, inspection, calibration, and maintenance of the contamination control program.

COL 11.2(3) The COL applicant is to determine whether contaminated laundry is sent to an offsite facility for cleaning or for disposal.

COL 11.2(4) The COL applicant is to prepare and provide the P&IDs.

COL 11.2(5) The COL applicant is to perform a site-specific cost-benefit analysis following the guidance in the regulatory requirements of NRC RG 1.110.

COL 11.2(6) The COL applicant is to provide reasonable assurance that the mobile or temporary equipment and interconnections to plant systems conform with the regulatory requirements and guidance of 10 CFR 50.34a, 10 CFR 20.1406, NRC RG 1.143, NRC RG 4.21, and ANSI/ANS 40.37. The COL applicant is to ensure that the implementation of mobile and temporary equipment meets the requirements of NUREG-0800 Sections 12.3 and 12.4 and that any impacts to the radiation zone maps are evaluated.

COL 11.2(7) The COL applicant is to develop a plant-wide NRC RG 4.21 Program following the guidance in NEI 08-08A for contamination control.

COL 11.2(8) The COL applicant is to maintain the complete documentation of system design, construction, design modifications, field changes, and operations and make them available for decommissioning planning and implementation.

COL 11.2(9) The COL applicant is to develop the procedure for the collection and shipment of mixed wastes, if and when they are generated, for offsite treatment. The generation of mixed liquid wastes is minimized by process control and the controlled use of hazardous chemicals.

COL 11.2(10) The COL applicant is to develop the interface design and provide the site-specific information for the LWMS effluent discharge, including radioactive release points, effluent temperature, the design (type, shape, and size) of flow orifices, and the sampling requirements following the guidance of NRC RG 1.21 and RG 4.15 and the standards incorporated therein by reference.

COL 11.2(11) The COL applicant is to prepare the site process control program and the site radiological environmental monitoring program.

COL 11.2(12) The COL applicant is to confirm the assumed dilution flow rate provided by cooling tower blowdown, dilution pump, or other plant discharges at the discharge point based on site-specific parameters.

COL 11.2(13) The COL applicant is to calculate dose to members of the public following the guidance of NRC RG 1.109 and NRC RG 1.113 using site-specific parameters and to compare the doses due to the liquid effluents with the numerical design objectives of Appendix I to 10 CFR 50, 10 CFR 20.1302, and 40 CFR 190.

COL 11.2(14) The COL applicant is to perform an analysis to demonstrate that the potential groundwater or surface water contamination concentrations resulting from radioactive release from the liquid-containing tank failure, are in compliance with the limits in 10 CFR 20, Appendix B, Table 2.

COL 11.3(1) The COL applicant is to prepare and implement the epoxy inspection and maintenance program in the GRS.

COL 11.3(2) The COL applicant is to perform a site-specific cost-benefit analysis following the guidance in NRC RG 1.110 for conformance with 10 CFR 50 Appendix I.

1.8-28 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (25 of 39)

Item No. Description COL 11.3(3) The COL applicant is to prepare and provide the piping and instrumentation diagram (P&ID) for the combined operating license application.

COL 11.3(4) The COL applicant is to prepare the operational procedures and maintenance programs related to leak detection and contamination control.

COL 11.3(5) The COL applicant is to maintain complete documentation of system design, construction, design modifications, field changes, and operations for decommissioning planning.

COL 11.3(6) The COL applicant is to prepare the site process control program and the site radiological environmental monitoring program.

COL 11.3(7) The COL applicant is also to perform the dose calculation using the total gaseous effluents from the site for comparison with the requirements of 40 CFR 190.

COL 11.3(8) The COL applicant is to perform an analysis using site-specific meteorological data to demonstrate that the potential airborne concentration resulting from GRS failure meets the requirements of 10 CFR 20, Appendix B, Table 2.

COL 11.3(9) The COL applicant is to prepare an ODCM following the guidance in NEI 07-09A template.

COL 11.4(1) The COL applicant can incorporate an onsite laundry facility for processing of contaminated clothing.

COL 11.4(2) The COL applicant is to perform a site-specific cost-benefit analysis following the guidance in NRC RG 1.110.

COL 11.4(3) The COL applicant is to provide reasonable assurance that the provisions and requirements of ANSI/ANS-40.37-2009 are met. The COL applicant is to provide reasonable assurance that mobile and temporary solid radwaste processing operations and its interconnection to plant systems conform with regulatory requirements and guidance such as 10 CFR 50.34a, 10 CFR 20.1406, NRC RG 1.143, IE Bulletin 80-10, and NRC RG 4.21. The COL applicant is to ensure that the implementation of mobile and temporary equipment meets the requirements of NUREG-0800 Sections 12.3 and 12.4 and that any impacts to the radiation zone maps are evaluated. The COL applicant is to prepare a plan to develop and use operating procedures and portable radiation monitoring instruments, so that regulatory requirements and guidance are followed.

COL 11.4(4) The COL applicant is to provide P&IDs.

COL 11.4(5) The COL applicant is to prepare the site process control program, which is consistent with the guidance of NEI 07-10A, and the site radiological environmental monitoring program.

COL 11.4(6) The COL applicant is to determine the number of shipment. If the generated waste exceeds the shipment amounts, the COL applicant is to provide details for how the extra generated waste that exceeds the shipment amounts will be handled.

COL 11.4(7) The COL applicant is responsible for the collection, temporary storage, and shipment of mixed waste for offsite treatment and disposal.

COL 11.4(8) The COL applicant is responsible for the provision of a site-wide IRSF for interim storage of radioactive wastes. The COL applicant is to provide reasonable assurance that the design and operation of such a facility meets the regulations and guidance specified in NUREG-0800, Section 12.3 and 12.4, including 10 CFR 20.1101, 10 CFR 20.1406, RG 4.21 and RG 8.8 COL 11.4(9) The COL applicant is to provide a mobile crane to retrieve a waste package that becomes stuck in the lifted condition or that is dropped.

COL 11.4(10) The COL applicant is also to provide operational procedures to properly ship low-level wastes to external sites in accordance with US NRC and US Department of Transportation (DOT) regulations.

1.8-29 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (26 of 39)

Item No. Description COL 11.4(11) The COL applicant is to prepare the operational procedures and maintenance programs for the SWMS as related to leak detection and contamination control.

COL 11.4(12) The COL applicant is to develop plant-wide RG 4.21 life-cycle planning for minimization of contamination program following the guidance in NEI 08-08A, in which the SWMS procedures and programs are to be integrated.

COL 11.4(13) The COL applicant is to maintain the complete documentation of system design, construction, design modifications, field changes, and operations.

COL 11.5(1) The COL applicant is to determine the WARN and ALARM setpoints of the PERMSS based on the site-specific conditions and operational requirements.

COL 11.5(2) The COL applicant is to develop an annual report that specifies the quantity of each principal radionuclide released to unrestricted areas in liquid and gaseous effluents.

COL 11.5(3) The COL applicant is to provide site-specific procedures that conform with the numerical guides of 10 CFR 50.34a and 10 CFR Part 50, Appendix I.

COL 11.5(4) The COL applicant is to prepare an offsite dose calculation manual (ODCM) that contains a description of the methodology and parameters for the calculation of the offsite doses for the gaseous and liquid effluents.

COL 11.5(5) The COL applicant is to provide analytical procedures and sensitivity for selected radio-analytical methods and types of sampling media for site-specific applications.

COL 11.5(6) The COL applicant is also to develop operational procedures in accordance with NRC RG 1.33 and NRC RG 4.15.

COL 11.5(7) The COL applicant is to develop a radiological and environmental monitoring program (REMP) in accordance with NUREG-1301 and NUREG-0133, and NRC RG 4.1, which describes the scope of the program, taking into account local and land use census data in identifying all potential radiation exposure pathways, associated radioactive materials present in liquid and gaseous effluent, and direct external radiation from SSCs.

COL 11.5(8) The COL applicant is to develop detailed locations, tubing installations, and provide the sampling method including the sampling frequency and time to acquire representative sampling.

COL 11.5(9) The COL applicant is to determine the safety class and installed location of the RE-165, RE-166, RE-167, and RE-168.

COL 11.5(10) The COL applicant is to provide operational procedures and maintenance programs related to leak detection and contamination control.

COL 11.5(11) The COL applicant is to design the sample nozzle location, sample line size, line routing/configuration/length, and monitor location to minimize the line loss in accordance with ANSI/HPS N13.1.

COL 12.1(1) The COL applicant is to provide the organizational structure to effectively implement the radiation protection policy, training, and reviews consistent with operational and maintenance requirements, while satisfying the applicable regulations and Regulatory Guides including NRC RGs 1.33, 1.8, 8.8, and 8.10.

COL 12.1(2) The COL applicant is to describe the operational radiation protection program to provide reasonable assurance that occupational and public radiation exposures are ALARA.

COL 12.1(3) The COL applicant is to describe how the plant follows the guidance provided in NRC RGs 8.2, 8.4, 8.7, 8.9, 8.13, 8.15, 8.20, 8.25, 8.26, 8.27, 8.28, 8.29, 8.34, 8.35, 8.36, and 8.38.

1.8-30 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (27 of 39)

Item No. Description COL 12.2(1) The COL applicant is to provide any additional contained radiation sources, such as instrument calibration radiation sources, that are not identified in Subsection 12.2.1.

COL 12.3(1) The COL applicant is to determine the areas that will require either electro or mechanical polishing.

COL 12.3(2) The COL applicant is to establish how the water chemistry pH control reduces radiation fields.

COL 12.3(3) The COL applicant is to provide the material composition and shielding properties of the following doors/hatches, and these thicknesses equivalent to the minimum required concrete shield thicknesses.

- Personnel Air Lock between Containment Annulus Area (100-C01) and Personnel Air Lock Entrance (100-A14A)

- Personnel Air Lock between Operating Area (156-C01) and Containment Entrance Area (156-A04B)

- Equipment Hatch between Operating Area (156-C01) and Equipment Hatch Access Room (156-A10A)

- Door between Equipment Hatch Access Room (156-A10A) and the building exterior

- Doors between Truck Bay (100-P08) and the building exterior Also, the COL applicant is to provide the service life of these doors/hatches and perform periodic in-service inspection and maintenance for these doors/hatches to provide reasonable assurance of functionality throughout the life of the plant.

COL 12.3(4) COL applicant is to provide information to ensure that radiation levels at the site boundary not exceed the limits of 40 CFR Part 190, from all radiation sources, including the outdoor tanks.

COL 12.3(5) The COL applicant is to provide portable instruments and the associated training and procedures in accordance with 10 CFR 50.34(f)(2)(xxvii) and the criteria in Item III.D.3.3 of NUREG-0737 as well as the guidelines of RG 8.8.

COL 12.3(6) The COL applicant is to determine the ARM setpoints for WARN, ALARM, and the containment purge isolation and fuel handling area emergency ventilation actuation signals, based on the site-specific conditions and operational requirements.

COL 12.3(7) The COL applicant is to determine the purpose and use of the room (063-P73). The access control for this room shall be changed accordingly in compliance with the guidance in RG 8.38. In addition, COL applicant is to specify any necessary radiation monitoring requirements, and any additional necessary design features and controls to ensure compliance with applicable regulations, including 10 CFR Part 20 and 10 CFR Part 36.

COL 12.4(1) The COL applicant is to estimate construction worker doses based on site-specific information, such as the number of operating units, distances from radiation sources, meteorological conditions, and construction schedule.

COL 12.4(2) The COL applicant is to provide operational procedures and programs including the development of a site radiological environmental monitoring program, to implement the minimization of contamination approach in accordance with NRC RG 4.21 and RG 4.22, as applicable, and the documentation required by 10 CFR 20.1501.

COL 12.4(3) The COL applicant is to implement concrete tunnels for piping of the systems that may include underground piping carrying contaminated or potentially contaminated fluid to minimize buried piping. The tunnels are coated with epoxy and are equipped with sumps with liquid detection level switches. If liquid is accumulated to the detectable level, an alarm is initiated in the MCR for operator actions.

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APR1400 DCD TIER 2 Table 1.8-2 (28 of 39)

Item No. Description COL 12.4(4) The COL applicant is to provide a respiratory protection program to minimize airborne radiological hazards while performing post-accident vital functions. The respiratory protection program should include the provisions of the positive pressure self-contained breathing apparatus (SCBA) with a minimum rated service life of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in the control room and air supply systems in areas where post-accident mission times may exceed the 1-hour SCBA supply and where necessary to account for uncertainty in respirator service life in performing post-accident missions. The COL applicant will also assess if replenishing the respirators during vital missions will result in any increase to the vital area mission times and doses.

COL 12.5(1) The COL applicant is to provide the operational radiation protection program, including the items described in Section 12.5.

COL 13.1(1) The COL applicant is to provide a description of the corporate or home office organization, its functions and responsibilities, and the number and the qualifications of personnel. The COL applicant is to be directed to activities such as the facility design, design review, design approval, construction management, testing, and operation of the plant.

COL 13.1(2) The COL applicant is to develop a description of experience in the design, construction, and operation of nuclear power plants and experience in activities of similar scope and complexity.

COL 13.1(3) The COL applicant is to describe its management, engineering, and technical support organizations. The description includes organizational charts for the current headquarters and engineering structure and any planned modifications and additions to those organizations to reflect the added functional responsibilities with the nuclear power plant.

COL 13.1(4) The COL applicant is to develop a description of the organizational arrangement. The description is to include organizational charts reflecting the current headquarters and engineering structure and any planned modifications and additions to reflect the added functional responsibilities associated with the addition of the nuclear plant to the applicants power generation capacity. The description shows how these responsibilities are delegated and assigned or expected to be assigned to each of the working or performance-level organizational units identified to implement these responsibilities. The description includes organizational charts reflecting the current corporate structure and the working- or performance-level organizational units that provide technical support for the operation.

COL 13.1(5) The COL applicant is to develop the description of the general qualifications in terms of educational background and experience for positions or classes of positions described in the organizational arrangement.

COL 13.1(6) The COL applicant is to develop a description of the structure, functions, and responsibilities of the onsite organization established to operate and maintain the plant.

COL 13.1(7) The COL applicant is to provide an organizational chart showing the title of each position, minimum number of persons to be assigned to duplicate positions, number of operating shift crews, and positions that require reactor operator and senior reactor operator licenses.

COL 13.1(8) The COL applicant is to provide organizational information such as the functions, responsibilities, and authorities of the plant position. The COL applicant is to develop a description of the line of succession of authority and responsibility for overall station operation in the event of unexpected temporary contingencies, and the delegation of authority.

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APR1400 DCD TIER 2 Table 1.8-2 (29 of 39)

Item No. Description COL 13.1(9) The COL applicant is to develop a description of the position titles, applicable operator licensing requirements for each, and the minimum numbers of personnel planned for each shift for all combinations of units proposed to be at the station in either operating or cold shutdown mode. The COL applicant is also to develop the description of shift crew staffing plans unique to refueling operations.

COL 13.1(10) The COL applicant is to provide a description of the education, training, and experience requirements for each management, operating, technical, and maintenance position in the operating organization.

COL 13.1(11) The COL applicant is to provide the qualification requirements of the initial appointees to plant positions for key plant managerial and supervisory personnel through the shift supervisory level.

COL 13.2(1) The COL applicant is to develop the description and schedule of the training program for licensed reactor operators and non-licensed plant staff.

COL 13.2(2) The COL applicant is to develop the site-specific training program by using NEI 06-13A as the template for the basic structure and content.

COL 13.2(3) The COL applicant is to provide a licensed plant staff training program in accordance with NEI 06-13A.

COL 13.2(4) The COL applicant is to provide a non-licensed plant staff training program in accordance with NEI 06-13A.

COL 13.2(5) The COL applicant is to develop training programs. The programs are to include a chart that shows the schedule of each part of the training program for each functional group of employees in the organization in relation to the schedule for preoperational testing, expected fuel loading, and expected time for examinations prior to plant criticality for licensed operators.

COL 13.2(6) The COL applicant is to determine the extent of the NRC guidance that is applicable to the facility training program or the justification of exceptions.

COL 13.3(1) The COL applicant is to develop the interfaces of design features with site-specific designs and site parameters.

COL 13.3(2) The COL applicant is to develop a comprehensive emergency plan. The plan is developed as a physically separate document and includes copies of letters of agreement (or other certifications) from state and local governmental agencies with emergency planning responsibilities.

COL 13.3(3) The COL applicant is to address an emergency classification and action level scheme as required by 10 CFR 50.47(b)(4).

COL 13.3(4) The COL applicant is to develop the security-related aspects of an emergency plan.

COL 13.3(5) The COL applicant is to develop a multi-unit site interface plan depending on the location of the new reactor on or near an operating reactor site with an existing emergency plan.

COL 13.3(6) The COL applicant is to develop emergency planning inspections, tests, analyses, and acceptance criteria.

COL 13.4(1) The COL applicant is to develop operational programs and provide schedules for implementation of the programs, as defined in SECY-05-0197. The COL applicant is to provide commitments for the implementation of operational programs that are required by regulation. In some instances, the programs may be implemented in phases, where practical, and the applicant is to include the phased implementation milestones.

1.8-33 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (30 of 39)

Item No. Description COL 13.4(2) The COL applicant is responsible for developing a leakage monitoring and prevention program for the systems, as specified in Subsection 5.5.2 in Chapter 16, Technical Specifications. The leakage monitoring and prevention program is to provide suitable methods and acceptance criteria as defined in NUREG-0737, Item III.D.1.1.

COL 13.4(3) The COL applicant is to develop an implementation plan for an inspection and monitoring program of the cladding material integrity of SG channel heads. The COL applicant is to provide a commitment for the implementation plan of the inspection and monitoring program.

COL 13.5(1) The COL applicant is to describe the administrative and operating procedures.

Administrative procedures provide for administrative control over safety-related activities for the operation of the facility. Operating procedures are used to ensure that routine operating, off-normal, and emergency activities are conducted in a safe manner. The COL applicant is to provide a description of the nature, content, and development process for the administrative and operating procedures, including preliminary schedules for preparation and target dates for completion (Reference 1 through 3).

COL 13.5(2) The COL applicant is to provide a program for developing and implementing administrative procedures.

COL 13.5(3) The COL applicant is to describe the different classifications of procedures the operators use in the MCR and locally in the plant for plant operations. The COL applicant is to describe the operating procedures that will be used by the operating organization (plant staff) to ensure that routine operating, off-normal, and emergency activities are conducted in a safe manner. The COL applicant is to identify the group within the operating organization responsible for maintaining the procedures and describe the general format and content of the different classifications.

COL 13.5(4) The COL applicant is to provide a program for developing and implementing operating procedures.

COL 13.5(5) The COL applicant is to provide a program for developing and implementing emergency operating procedures.

COL 13.5(6) The COL applicant is to describe the procedures that provide coverage for other safety-related plant operating activities (i.e., operating activities not procedurally covered under the operating or emergency operating procedure programs), including related maintenance activities. The COL applicant is to provide a description of the nature, content, and development process for the maintenance and other operating procedures, including preliminary schedules for preparation and target dates for completion. In addition, the COL applicant is to describe how these procedures are classified, describe the general format and content of the various classifications, and identify the group(s) within the operating organization responsible for performing and maintaining the procedures.

COL 13.5(7) The COL applicant is to provide a program for developing and implementing procedures that provide coverage for other safety-related plant operating activities (i.e., operating activities not procedurally covered under the operating or emergency operating procedure programs),

including related maintenance activities.

COL 13.5(8) The COL applicant is to provide a program for developing shutdown procedure including the installation and removal order of the pressurizer manway and the nozzle dam.

COL 13.7(1) The COL applicant is to develop the description of the fitness-for-duty programs during construction and for the operating plant.

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APR1400 DCD TIER 2 Table 1.8-2 (31 of 39)

Item No. Description COL 14.2(1) The COL applicant is to develop the site-specific organization and staffing level appropriate for its facility to implement the initial test program. The COLs plant operating and plant technical staff should participate, to the extent practical, in developing and conducting the Initial Test Program and evaluating the test results.

COL 14.2(2) The COL applicant is to prepare the site-specific preoperational and startup test specifications and test procedures and/or guidelines that is to be used for the conduct of the plant Initial Test Program. The preoperational and startup test procedures should have controls in place to ensure that test procedures include appropriate prerequisites, objectives, safety precautions, initial test conditions, methods to direct and control test performance and test acceptance criteria by which the test is evaluated. Testing performed at other than design operating conditions for systems is to be reconciled either through the test acceptance criteria or post-test data analysis. These procedures are to be submitted at least 60 days prior to their intended use to the NRC staff for review as described in Subsection 14.2.11.

COL 14.2(3) The COL applicant is to prepare a startup administrative manual (SAM) which contains administrative controls that govern the conduct of each major phase of the ITP. This description should include the administrative controls used to ensure that necessary prerequisites are satisfied for each major phase and for individual tests. The COL applicant should also describe the methods to be followed in initiating plant modifications or maintenance tasks that are deemed to be necessary to conduct the ITP. This description should include methods used to ensure retesting following such modifications or maintenance. In addition, the description should discuss the involvement of design organizations with the COL applicant in reviewing and approving proposed plant modifications. The COL applicant should also describe in the SAM adherence to approved test procedures during the conduct of the ITP as well as the methods for effecting changes to approved test procedures.

COL 14.2(4) The COL applicant is to develop the test procedure including a listing of the high- and moderate-energy piping systems inside containment that are covered by the vibration, thermal expansion, and dynamic effects testing program.

COL 14.2(5) The COL applicant is to develop the test procedure including a listing of the different flow modes to which the systems will be subjected during the vibration, thermal expansion, and dynamic effects testing program to confirm that the piping systems, restraints, components, and supports have been adequately designed to withstand flow-induced dynamic loadings under the steady-state and operational transient conditions anticipated during service.

COL 14.2(6) The COL applicant is to develop the test procedure including a description of the thermal motion monitoring program for verification of snubber movement, adequate clearances and gaps, the acceptance criteria, and the method regarding how motion will be measured.

COL 14.2(7) The COL applicant is to perform review and evaluation of individual test results in a test report made available to NRC personnel after preoperational and startup tests are completed.

The specific test acceptance criteria for determining success or failure of a test shall be included in the test report approval of the test results. The test report should also include test results associated with any license conditions in the plant specific Initial Test Program.

COL 14.2(8) The COL applicant is responsible for establishing hold points at selected milestones throughout the power ascension test phase to ensure that designated personnel or groups evaluate and approve relevant test results before proceeding to the next power ascension test phase. At a minimum, the COL applicant should establish hold points at approximately 25-percent, 50- percent, and 75-percent power-level test conditions for pressurized-water reactors.

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APR1400 DCD TIER 2 Table 1.8-2 (32 of 39)

Item No. Description COL 14.2(9) The COL applicant is responsible for retaining preoperational and startup test procedures and test results as part of the plants historical records in accordance with 10 CFR 50.36, Technical Specification, 10 CFR 50.71, Maintenance of Records, Making of Reports, 10 CFR 50, Appendix B, Criterion XVII, Test Records, and RG 1.28, Quality Assurance Program Criteria (Design and Construction).

COL 14.2(10) The COL applicant is to describe its program for reviewing available information on reactor operating and testing experiences and discusses how it used this information in developing the initial test program. The description is to include the sources and types of information reviewed, the conclusions or findings, and the effect of the review on the initial test program.

COL 14.2(11) The COL applicant is to provide a schedule for the development of plant procedures, as well as a description of how, and to what extent, the plant operating, emergency, and surveillance procedures are use-tested during the initial test program.

COL 14.2(12) The COL applicant that references the APR1400 design certification is to identify the specific operator training to be conducted as part of the low-power testing program related to the resolution of TMI Action Plan Item I.G.1, as described in (1) NUREG-0660 - NRC Action Plans Developed as a Result of the TMI-2 Accident, Revision 1, August 1980 and (2)

NUREG-0737 - Clarification of TMI Action Plan Requirements.

COL 14.2(13) The COL applicant is to develop a sequence and schedule for the development of the plant operating and emergency procedures should allow sufficient time for trial use of these procedures during the Initial Test Program. The sequence and schedule for plant startup is to be developed by the COL applicant to allow sufficient time to systematically perform the required testing in each phase.

COL 14.2(14) The COL applicant is to perform the appropriate interface testing of the PERMSS and ARMS monitors with ERDS.

COL 14.2(15) The COL applicant is to prepare the preoperational test of cooling tower and associated auxiliaries, and raw water and service water cooling systems.

COL 14.2(16) The COL applicant is to develop the test program of personnel monitors, radiation survey instruments, and laboratory equipment used to analyze or measure radiation levels and radioactivity concentrations.

COL 14.2(17) The COL applicant is to prepare the site-specific preoperational and startup test specification and test procedure and/or guideline for offsite communication system.

COL 14.2(18) The COL applicant is to prepare the pre-operational test of ultimate heat sink pump house.

COL 14.2(19) The COL applicant is to prepare the testing and verification of ultimate heat sink cooling chains.

COL 14.3(1) The COL applicant is to provide the ITAAC for the site-specific portion of the plant systems specified in Subsection 14.3.3.

COL 14.3(2) The COL applicant is to provide a design ITAAC closure schedule for implementing the V&V design ITAAC as addressed in Subsection 14.3.2.9.

COL 14.3(3) The COL applicant is to provide the proposed ITAAC for the facilitys emergency planning not addressed in the DCD in accordance with RG 1.206.

COL 14.3(4) The COL applicant is to provide the proposed ITAAC for the site specific facilitys physical security hardware not addressed in the DCD in accordance with RG 1.206.

1.8-36 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (33 of 39)

Item No. Description COL 15.0(1) The COL applicant is to perform the radiological consequence analysis using site-specific

/Q values, unless the /Q values used in the DCD envelop the site-specific short-term or long-term /Q values of the DCD, and to show that the resultant doses are within the guideline values of 10 CFR 50.34 for EAB and LPZ and that of 10 CFR Part 50, Appendix A, GDC 19 for the MCR and TSC.

COL 15.0(2) The COL applicant is to perform the radiological consequence analysis and demonstrate that the related dose limits specified in 10 CFR Part 50, Appendix A GDC 19 are not exceeded, if the interval value of re-opening the closed outside air intake isolation dampers is less than that specified in Subsection 15.0.3.5.

COL 15.0(3) The COL applicant is to perform the radiological consequence analysis and demonstrate that the related dose limits specified in 10 CFR Part 50, Appendix A GDC 19 are not exceeded, if the intake damper re-opening and closing time exceeds that specified in Subsection 15.0.3.5 or there are any other aspects of the design of the dampers that are non-conservative compared to what is used in Subsection 15.0.3.5.

COL 16.1(1) The choice of units is a COL information item to be resolved by the COL applicant COL 16-3(1) Applicability of Reviewers Notes in Bases Sections B3.0(LCO3.0.9), B3.6.7, B3.7.5, B3.7.11, B3.7.12, B3.8.1, and B3.9.2 are to be determined by the COL applicant.

COL 16-3.0(1) The COL applicant is to prepare and submit for NRC staff review a site specific risk evaluation supporting the adoption of LCO 3.0.9 that is based on the site specific probabilistic risk assessment as described in FSAR chapter 19.

COL 16-3.6(1) The COL applicant is to provide the minimum number of bolt, completion time, and surveillance frequency for shutdown operations. The value will be determined based on plant specific Shutdown Evaluation Report to satisfy the 10 CFR 50.34 dose limits.

COL 16-3.7(1) The COL applicant is to provide the completion time for required action for condition A&C or provide a risk-informed justification to increase the completion time.

COL 16-3.7(2) The COL applicant is to provide the completion time, and surveillance frequency for ultimate heat sink. Ultimate heat sink design value varies depending on site characteristics.

COL 16-3.7(3) The COL applicant is to provide the details of specific toxic chemicals of mobile and stationary sources and evaluate the MCR habitability based on the recommendation in RG 1.78. The specific toxic gas concentrations in the air intakes will vary depending on site.

If the applicant determines that the maximum concentrations in the MCR for a given toxic gases do not exceed the toxicity limits from RG 1.78, toxic gas detector and toxic gas isolation mode are not required and the bracketed phrases are deleted.

COL 16-3.8(1) The COL applicant is to provide the specific value in accordance with EDG manufactures specific design characteristics.

COL 16-3.8(2) The COL applicant is to determine plant operation MODES which allow the Surveillance depend on the plant operation and surveillance policy. The MODES restrictions may be deleted if the COL applicant demonstrates that the plant safety is maintained or enhanced when the surveillance is performed in restricted MODES.

COL 16-3.8(3) The COL applicant is to determine EDG power factor as applicable to the tests. EDG operation power factor depends on plant specific EDG Class 1E loads and offsite power condition.

COL 16-3.8(4) The COL applicant is to provide the specific value in accordance with EDG manufacturers specific design recommendation.

1.8-37 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (34 of 39)

Item No. Description COL 16-3.9(1) The COL applicant is to provide the minimum number of bolt for equipment hatch. The value will be determined based on plant specific Shutdown Evaluation Report to satisfy the 10 CFR 50.34 dose limits.

COL 16-4.1(1) Information on site location is to be provided by the COL applicant.

COL 16-5(1) Applicability of Reviewers Notes in Tech Spec 5.1, 5.2.2, 5.3, 5.5.3, 5.5.11, 5.5.17, 5.5.19, and 5.6.4 are to be determined by the COL applicant.

COL 16-5.3(1) The requirement for unit staff qualification shall be determined by COL applicant based on latest NRC RG 1.8 and ANSI standard acceptable to NRC staff.

COL 16-5.4(1) The COL applicant will determine the modification of core protection calculator (CPC) addressable constants based on plant specific data.

COL 16-5.5(1) The methodology for gaseous radioactivity quantities and the liquid radwaste quantities is to be provided by the COL applicant.

COL 16-5.5(2) The FSAR reference on setpoint control document is to be specified by the COL applicant.

COL 16-5.6(1) A single submittal of reporting on multiple unit stations is to be determined in COL stage.

COL 16-5.6(2) The COL applicant will determine the format of the table. Either format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979, or use another format that is acceptable to NRC.

COL 17.1(1) The COL applicant is to establish and implement a QA program that is applicable to site-specific design activities during the plant construction and operation phases.

COL 17.2(1) The COL applicant establishes and implements a QA program that is applicable to site-specific design activities during the plant construction and operation phases.

COL 17.3(1) The COL applicant is to establish and implement a QA program that is applicable to site-specific design activities during the plant construction and operation phases.

COL 17.4(1) The APR1400 application will update the description of the D-RAP to include relevant site-and plant-specific information (e.g., design, program, procedural, and organizational information). This includes identifying the SSCs within the scope of the plant-specific RAP (i.e., the RAP SSCs identified in the DC, updated using COL site- and plant-specific information) and establishing the programmatic controls of D-RAP to be applied during the COL design and construction activities prior to initial fuel load.

COL 17.4(2) The APR1400 application will specify appropriate QA controls for the non-safety-related RAP SSCs in accordance with the provisions in Part V, Non-safety-Related SSC Quality Controls, of SRP Section 17.5. This includes providing corrective actions for potential design and pre-operational errors that could degrade non-safety-related RAP SSCs.

COL 17.4(3) The APR1400 application will propose a process for integrating the RAP into operational programs (e.g., maintenance rule program, QA program, inservice inspection, inservice testing, surveillance testing, and maintenance programs). The process should also address the (1) establishment of reliability, availability, or condition performance goals for the RAP SSCs, (2) establishment of performance and condition monitoring requirements to provide reasonable assurance that RAP SSCs do not degrade to an unacceptable level of reliability, availability, or condition during plant operations, (3) for non-safety-related RAP SSCs, establishment of QA controls for activities during the operations phase in accordance with the provisions in Part V of SRP Section 17.5, and (4) consideration of dominant failure modes of RAP SSCs in meeting the objectives of the RAP during plant operation.

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Item No. Description COL 17.5(1) The COL applicant is to establish and implement a QA program that is applicable to site-specific design activities related to the plant construction and operation phases.

COL 17.6(1) The COL applicant is to provide in its Final Safety Analysis Report a description of the Maintenance Rule program and a plan for implementing it to meet the requirements of 10 CFR 50.65.

COL 18.12(1) The COL applicant is to provide the human performance monitoring program.

COL 19.0(1) The COL applicant is either to confirm that the PRA in the design certification bounds the site-specific design information and any design changes or departures, or to update the PRA to reflect the site-specific design information and any design changes or departures.

COL 19.1(1) The COL applicant is to describe the uses of PRA in support of licensee programs, and to identify and describe risk-informed applications being implemented during the combined license application phase.

COL 19.1(2) The COL applicant is to describe the uses of PRA in support of licensee programs, and identify and describe risk-informed applications being implemented during the construction phase.

COL 19.1(3) The COL applicant is to describe the uses of PRA in support of licensee programs, and identify and describe risk-informed applications being implemented during the operational phase.

COL 19.1(4) The COL applicant is to review as-designed and as-built information and conduct walkdowns as necessary to confirm that the assumptions used in the PRA (including PRA inputs to RAP and SAMDA) remain valid with respect to internal events, internal flood and fire events (fire barrier and fire barrier penetrations, routings and locations of pipe, cable, and conduit), and HRA analyses (development of operating procedures, emergency operating procedures, and severe accident management guidelines and training), external events including PRA-based seismic margins and HCLPF fragilities, and LPSD procedures.

COL 19.1(5) The COL applicant is to conduct a peer review of the PRA relative to the industry PRA Standard prior to use of the PRA to support risk-informed applications, as applicable.

COL 19.1(6) The COL applicant is to describe the PRA maintenance and upgrade program.

COL 19.1(7) The COL applicant is to develop management procedures for charging pump operation, following recovery from a loss of offsite power (LOOP), to ensure that deboration is not resumed until after at least one Reactor Coolant Pump (RCP) has been restarted.

COL 19.1(8) The COL applicant is to confirm that the PRA-based seismic margin assessment is bounding for the selected site, and to update the assessment to include site-specific SSC and soil effects (including sliding, overturning liquefaction, and slope failure). The COL applicant is to confirm that the as-built plant has adequate seismic margin and do not exceed the CDF and LERF design targets specified in Subsection 1.2.1.1.1 e.

The COL applicant is to demonstrate that HCLPF capacity is equal to or exceed 1.67 times the GMRS for site-specific structures (ESWIS and CCW Hx Building).

The COL applicant is to demonstrate that HCLPF capacity is equal to or exceed 1.67 times the CSDRS for BOP components and is to complete the SEL.

The COL applicant is to demonstrate that the seismic capacity for equipment qualified by testing should remain functionally operational within 1.67 times the required response spectra (CSDRS-based RRS) in the procurement specification.

1.8-39 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (36 of 39)

Item No. Description COL 19.1(9) When developing post-earthquake safe shutdown procedures, the COL applicant should consider the potential for multiple spurious alarms from photoelectric detectors following a seismic event.

COL 19.1(10) The COL applicant needs to ensure that screened events do not have a site- specific susceptibility and do not exceed the CDF and LERF design targets specified in Subsection 1.2.1.1.1 e. The COL applicant is to address the following issues with a site-specific risk assessment, as applicable: dam failure, external flooding, extreme winds and tornadoes, industrial or military facility, pipeline accident, release of chemicals from onsite storage, river diversion, sandstorm, toxic gas, transportation accidents, and aircraft crash event.

In addition, the COL applicant is to ensure the site-specific susceptibility is not an outlier for the following issues, as applicable: avalanche, biological events, coastal erosion, dam failure, drought, forest fire, high summer temperature, landslide, low lake/river water level, low winter temperature, and volcanic activity. The COL applicant is to ensure operation of the emergency disel generator sequencer throughout low power and shutdown operations (not including defueled plant operating states).

COL 19.1(11) The COL applicant is to develop outage management procedures that limit planned maintenance that can potentially impair one or both SC trains during the shutdown modes.

COL 19.1(12) The COL applicant is to develop procedures and a configuration management strategy to address the period of time when one SC train is unexpectedly unavailable (including the termination of any testing or maintenance that can affect the remaining train and restoration of all equipment to its nominal availability).

COL 19.1(13) The COL applicant is to establish procedures for closing the containment hatch (after being opened during during LPSD operations) to promptly re-establish the containment as a barrier to fission product release. This guidance must include steps that allow for sealing of the hatch with four bolts (versus the 40 bolts used to secure the hatch during at-power operation); four bolts are sufficient to secure the hatch so that no visible gap can be seen between the seals and the sealing surface.

COL 19.1(14) The COL applicant is to develop procedures specifying that a fire watch be present when hot work is being performed.

COL 19.1(15) The COL applicant is to develop the fire barrier management procedures that direct the appropriate use of a fire watch and use of the isolation devices with a quick-disconnect mechanism for hose and cables that bleach a fire barrier.

COL 19.1(16) The COL applicant is to develop procedures and operator training for reliance (during fire response) on undamaged instrumentation (when the location of the fire is known). The COL applicant is to ensure Alarm Response Procedures related to Loss of all RCP seal cooling including, but not limited to, partial and total loss of the component cooling water system, and partial and total loss of the essential service water system, include steps to trip the reactor followed by immediate trip of the RCPs if RCP seal cooling cannot be re-established within the time limit specified by the manufacturer.

COL 19.1(17) The COL applicant is to develop a configuration control program requiring that, during Modes 4, 5, and 6, the watertight flood doors and fire doors be maintained closed in at least one quadrant. Furthermore, the COL applicant is to incorporate, as part of the aforementioned configuration control program, a provision that if the flood or fire doors to this designated quadrant must be opened for reasons other than normal ingress/egress, a flood or fire watch must be established for the affected doors.

COL 19.1(18) The COL applicant is to describe the uses of PRA in support of licensee programs such as Maintenance Rule implementation during the operational phase.

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Item No. Description COL 19.1(19) The COL applicant is to describe the uses of PRA in support of licensee programs such as the reactor oversight process during the operational phase.

COL 19.1(20) The COL applicant is to perform the seismic-fire interactions walkdown to confirm a qualitative seismic-fire interaction assessment.

COL 19.1(21) The COL applicant is to develop outage procedures to ensure that in fire compartments containing post-seismic or post-fire safe shutdown equipment that: 1) the seismic ruggedness of temporary ignition sources is adequate, or that the duration that these temporary ignition sources are in these areas is minimized, 2) the seismic ruggedness of temporary equipment such as scaffolding in fire compartments containing potential seismic-fire ignition sources, or near fire protection equipment is adequate, and 3) either the duration of activities which could impact manual firefighting is minimized, or alternative firefighting equipment (e.g.,

pre-stage portable smoke removal equipment, prestage additional firefighting equipment, etc.) is supplied.

COL 19.1(22) The COL applicant is to demonstrate that failure of buildings that are not seismic Category I (e.g., turbine building and compound building) does not impact SSCs designed to be seismic Category I.

COL 19.1(23) The COL applicant is to ensure that asymmetric conditions due to modeling simplicity will be addressed or properly accounted for when the PRA is used for decision making.

COL 19.1(24) The COL applicant will demonstrate that maintenance-induced floods are negligible contributors to flood risk when the plant specific data are available.

COL 19.1(25) SAMGs are entered to initiate SI with the core exit thermocouple indicating 1200 °F.

COL 19.2(1) The COL applicant is to perform and submit site-specific equipment survivability assessment including flooding effect in accordance with 10 CFR 50.34(f) and 10 CFR 50.44 which reflects the equipment identified and the containment atmospheric assessments of temperature, pressure and radiation described in Subsection 19.2.3.3.7.

COL 19.2(2) The COL applicant will demonstrate that the covers for large penetrations such as equipment hatch and personnel airlocks meet the Service Level C requirements in Subsection NE-3220 of the ASME code and explain how the consideration of containment leakage is accounted for when modeling local regions of containment.

COL 19.2(3) The COL applicant and/or holder is to develop and submit an accident management plan including the evaluation of the effect of higher water level in the cavity on steam explosion loading when using In-Vessel Retention and External Reactor Vessel Cooling for accident management.

COL 19.3(1) The COL applicant is to perform site-specific seismic hazard evaluation and seismic risk evaluation as applicable in accordance with NTTF Recommendation 2.1 as outlined in the NRC RFI.

COL 19.3(2) The COL applicant is to address the flood requirements for wet sites COL 19.3(3) The COL applicant is to develop the details for offsite resources.

COL 19.3(4) The COL applicant is to address the details of selecting suitable storage locations for FLEX equipment that provide reasonable protection during specific external events as provided in NEI 12-06 guidance Sections 5 through 9, and the details of the guidance for storage of FLEX equipment provided in the Technical Report (Reference 5) Section 6.2.9.

COL 19.3(5) The COL applicant is to 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 Section 19.3 in support of BDBEE mitigation strategies.

1.8-41 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (38 of 39)

Item No. Description COL 19.3(6) The COL applicant is to confirm and ensure that the raw water tank 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).

COL 19.3(7) The COL applicant is to provide site-specific details of the electrical connection points (including locations, accessibility, voltage level, and electrical classification) for the mobile GTGs.

COL 19.3(8) The COL applicant is to ensure that all the connections (refer to Table 6-5 in Reference 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).

COL 19.3(9) The COL applicant is to address site-specific strategies to mitigate BDBEEs as specified in the NRC Order EA-12-049.

COL 19.3(10) The COL applicant is to propose a License Condition to verify the development and implementation of the guidance, strategies, and programs for the mitigation strategies for ensuring core cooling, containment function, and spent fuel pool cooling capabilities during an extended loss of ac power event at an APR1400 nuclear power plant.

COL 19.3(11) The COL applicant is to provide a description of the operational programs that will provide assurance of the functional capability of the pumps, valves, and dynamic restraints used in the mitigation strategies for ensuring core cooling, containment function, and spent fuel pool cooling capabilities during an extended loss of ac power event at an APR1400 nuclear power plant.

COL 19.3(12) The COL applicant is to address SFP level instrumentation maintenance procedure development and perform training as specified in NRC Order EA-12-051.

COL 19.3(13) The COL applicant is to address development of EOPs, SAMGs, and EDMGs that incorporate lessons learned from TEPCO's Fukushima Dai-Ichi nuclear power plant accident as addressed in SECY-12-0025.

COL 19.3(14) The COL applicant is to address enhancement of the offsite communication system as specified in the NRC Request for Information pertaining to NTTF Recommendation 9.3.

COL 19.3(15) The COL applicant is to address staffing for large-scale natural events as specified in the NRC RFI pertaining to NTTF Recommendation 9.3.

COL 19.4(1) The COL applicant is to provide provision and procedure/guidance for implementing SFP strategies, including the minimum pump discharge pressure required to support the minimum flow requirements. All portable equipment is to be stored at least 91.44 m (100 yards) from the auxiliary building. The COL applicant is to specify the external water source(s) to be used.

COL 19.4(2) The COL applicant is to provide fire fighting response strategy.

COL 19.4(3) The COL applicant is to provide procedure/guidance for implementing the IRWST makeup strategy.

COL 19.4(4) The COL applicant is to determine the portable DC power source and connection device to SG pressure and level indication panels.

1.8-42 Rev. 2

APR1400 DCD TIER 2 Table 1.8-2 (39 of 39)

Item No. Description COL 19.4(5) [Deleted]

COL 19.4(6) The COL applicant is to provide procedure/guidance for implementing the manually depressurize SGs to reduce inventory loss.

COL 19.4(7) The COL applicant is to provide procedure/guidance for implementing the manual operation of TDAFW pump.

COL 19.4(8) The COL applicant is to provide procedure/guidance for implementing the manually depressurize SGs and use portable pump.

COL 19.4(9) The COL applicant is to provide procedure/guidance for implementing the makeup auxiliary feedwater tank.

COL 19.4(10) The COL applicant is to provide procedure/guidance for implementing the flooding containment with portable pump.

COL 19.4(11) The COL applicant is to provide portable equipment including portable pump, hose and connections for phase 3 strategies.

COL 19.5(1) When the reactor head is untensioned and before the refueling pool is flooded up, administrative controls will be in place to ensure that no trains of SI and shutdown cooling, including the necessary power and cooling water support systems are out of service for maintenance.

1.8-43 Rev. 2

APR1400 DCD TIER 2 1.9 Conformance with Regulatory Criteria The conformance of the APR1400 design with U.S. Nuclear Regulatory Commission (NRC) regulatory criteria is documented in this section. Regulatory criteria include NRC Regulatory Guides (RGs), Standard Review Plans (SRPs), generic issues including Three Mile Island (TMI) requirements, operational experience (generic communications), and advanced and evolutionary light-water reactor design issues per Subsections C.I.1.9.1 through C.I.1.9.5 of NRC RG 1.206 (Reference 1).

In addition, the conformance with post-Fukushima NRC recommendations and requirements are addressed in section 1.9.6.

The combined license (COL) applicant is to address an evaluation of the conformance with regulatory criteria for the site-specific portion and operational aspects of the facility (COL 1.9(1)).

1.9.1 Conformance with Regulatory Guides This section provides an evaluation of conformance with the following groups of NRC RGs:

a. Division 1, Power Reactors

b. Division 4, Environmental and Siting

c. Division 5, Materials and Plant Protection

d. Division 8, Occupational Health Conformance with applicable active guides is summarized in Table 1.9-1. The evaluation includes an identification and description of deviations from the guidance in the NRC RGs, as well as suitable justifications for exceptions or any alternative approaches. For NRC RGs not applicable to design, the reason for non-applicability is specified in the Conformance or Summary Description of Deviation column of Table 1.9-1. Table 1.9-1 is also cross-referenced to applicable sections of the design control document (DCD).

NRC RGs that have been withdrawn or are not publicly available are excluded from Table 1.9-1.

1.9-1 Rev. 2

APR1400 DCD TIER 2 1.9.2 Conformance with Standard Review Plan In accordance with 10 CFR 52.47(a)(9) (Reference 2), this subsection provides the APR1400 conformance with the acceptance criteria for each section of the SRP in effect 6 months before the docket date of the application. The evaluation results are presented in Table 1.9-2. The evaluation includes the identification and description of deviations from the SRP. Where differences exist, specific sections are identified and further details relevant to each SRP deviation are addressed.

1.9.3 Generic Issues In accordance with 10 CFR 52.47(a)(21), this subsection addresses the proposed technical resolution for all unresolved safety issues (USIs) and medium-and high-priority generic safety issues (GSIs) that are identified in the version of NUREG-0933 (Reference 3),

current as of the date 6 months before the docket date of the application, and that are technically relevant to the design. USIs and GSIs applicable to the APR1400 design were identified using the criteria given in Regulatory Position C.IV.8 of NRC RG 1.206.

Appendix B to NUREG-0933 (Rev. 25) (Reference 4), issued on September 2011, which is current on the date up to 6 months before the docket date of application, was used to identify new generic issues applicable to the APR1400 design certification. The evaluation results for each issue with the cross-reference to related sections are provided in Table 1.9-3.

In accordance with the requirements of 10 CFR 52.47(a)(8), the evaluations were also performed to assess conformance with any technically relevant portions of the TMI requirements set forth in 10 CFR 50.34(f) (Reference 5), except paragraphs (f)(1)(xii),

(f)(2)(ix), and (f)(3)(v). Table 1.9-4 addresses the assessment results of conformance with the TMI requirements and is cross-referenced to related DCD sections. TMI requirements applicable to the reactor types other than pressurized water reactor (PWR) or a specific vendor design are excluded from Table 1.9-4.

1.9.4 Operational Experience (Generic Communications)

The requirements of 10 CFR 52.47(a)(22) and Regulatory Position C.I.1.9.3 of NRC RG 1.206 specify that design certification applicants are to provide information necessary to demonstrate how the plant design incorporates operating experience insights from Generic Letters (GLs) and bulletins issued after the most recent revision of the applicable SRP and 1.9-2 Rev. 2

APR1400 DCD TIER 2 6 months before the docket date of the application or demonstrate comparable international operating experience.

The APR1400 design is an evolutionary plant design that has been developed based on Combustion Engineerings System 80+ plant.

GLs and bulletins issued after the March 2007 revision of the SRP have been assessed to address how the applicable operating experience has been incorporated into the APR1400 design. Table 1.9-5 provides the applicability of the generic communications to the APR1400 design and the results of the assessment.

1.9.5 Advanced and Evolutionary Light-Water Reactor Design Issues Section C.I.1.9.5 of NRC RG 1.206 specifies that the applicant is to address the licensing and policy issues developed by the NRC and documented in the Office of the Secretary of the Commission (SECY) documents and the associated Staff Requirements Memoranda (SRM) for advanced and evolutionary light water reactor (LWR) designs that apply to the proposed facility design.

Table 1.9-6 provides a discussion of the applicability of the SECYs listed in Section C.I.1.9.5 of NRC RG 1.206 to the APR1400 design. Table 1.9-7 provides a discussion of individual issues specified in SECY-93-087 (Reference 6) and provides cross references to relevant to the DCD chapters, sections, or subsections.

1.9.6 Conformance with Post-Fukushima NRC Recommendations and Requirements As a result of the Fukushima Dai-Ichi event, additional requirements have been established to manage and mitigate external events that are beyond the design basis of the plant. This section addresses the APR1400 conformance with SECY-12-0025 including the requirements contained in NRC Orders EA-12-049 and EA 12-051 and the related request for information. The specific details of addressing the Tier 1, 2 and 3 NTTF items are discussed in DCD 19.3.

Table 1.9-8 titled, APR1400 Strategy for Addressing Tier 1, 2 and 3 NTTF Recommendations provides response summary to SECY-11-0093, SECY-11-0137, SECY-11-0025, NRC Orders EA-12-049 and EA-12-051. Additionally, this table also provides the reference section(s) in the APR1400 design (Technical Report No.

1.9-3 Rev. 2

APR1400 DCD TIER 2 APR1400-E-P-NR-14005-P), applicable DCD section and COL action required to specifically address the NTTF recommendations and requirements.

1.9.7 Part 21 Notification of Failure to Comply or Existence of a Defect and Its Evaluation Conformance with 10 CFR Part 21 is a necessary requirement in the APR1400 design control process. Assessment and management of the design or other issues against the reporting requirements is required during both the development and implementation of the design. At of the time of the ARP1400 design certification application submission, no issues related to the APR1400 design that meet the reporting criteria of 10 CFR Part 21 (Reference 7) had been identified.

1.9.8 Combined License Information COL 1.9(1) The COL applicant is to provide an evaluation of the conformance with the regulatory criteria for the site-specific portions and operational aspects of the facility.

1.9.9 References

1. Regulatory Guide 1.206, Combined License Applications for Nuclear Power Plants (LWR Edition), U.S. Nuclear Regulatory Commission, June 2007.

2. 10 CFR 52.47, Contents of Applications; Technical Information, U.S. Nuclear Regulatory Commission

3. NUREG-0933, Resolution of Generic safety Issues, Rev. 34, U.S. Nuclear Regulatory Commission, September 2011, (includes Supplements 1-34).

4. NUREG-0933, Appendix B, Applicability of NUREG-0933 Issues to Operating and Future Reactor Plants, Rev. 25, U.S. Nuclear Regulatory Commission, September 2011.

5. 10 CFR 50.34(f),Additional TMI-related Requirements, U.S. Nuclear Regulatory Commission 1.9-4 Rev. 2

APR1400 DCD TIER 2

6. SECY-93-087, Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs, U.S. Nuclear Regulatory Commission, July 1993

7. 10 CFR Part 21, Reporting of Defects and Noncompliance, U.S. Nuclear Regulatory Commission.

8. APR1400-F-C-NR-14001-P, CPC Setpoint Analysis Methodology for APR1400, Rev.2, KHNP, November 2017.

1.9-5 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (1 of 35)

APR1400 Conformance with Regulatory Guides Revision/

NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.1 Net Positive Suction Head for Emergency Core 11/1970 The APR1400 conforms with the regulatory position with Table 6.2.2-1 Cooling and Containment Heat Removal System the following exception:

Pumps

• Calculations of available NPSH for the emergency core cooling and containment heat removal pumps were performed assuming that the containment pressure during post-accident conditions is equal to the vapor pressure of the liquid in the containment. This assumption provides reasonable assurance that the actual available NPSH is always greater than the calculated available NPSH, which meets the intent of the regulatory position.

1.3 Assumptions Used for Evaluating the Potential Rev. 2 Not applicable (BWR) N/A Radiological Consequences of a LOCA for 06/1974 Boiling Water Reactors 1.4 Assumptions Used for Evaluating the Potential Rev. 2 The APR1400 applies NRC RG 1.183 instead of this N/A Radiological Consequences of a Loss of Coolant 06/1974 NRC RG.

Accident for Pressurized Water Reactors 1.5 Assumptions Used for Evaluating the Potential 03/1971 Not applicable (BWR) N/A Radiological Consequences of a Steam Line Break Accident for Boiling Water Reactors 1.6 Independence Between Redundant Standby (On- 03/1971 The APR1400 conforms with this NRC RG. 8.1.3.3, 8.3.1.2.2, Site) Power Sources and Between Their 8.3.2.2.2 Distribution Systems 1.7 Control of Combustible Gas Concentration in Rev. 3 The APR1400 conforms with this NRC RG. 6.1.1.1, 6.2.5.1 Containment 03/2007 1.9-6 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (2 of 35)

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.8 Qualification & Training of Personnel for Rev. 3 Not applicable (COL) N/A Nuclear Power Plants 05/2000 1.9 Application and Testing of Safety-Related Diesel Rev. 4 The APR1400 conforms with this NRC RG. 8.1.3.3, 8.3.1.2.2 Generators in Nuclear Power Plants 03/2007 1.11 Instrument Lines Penetrating Primary Reactor Rev. 1 The APR1400 conforms with this NRC RG. 3.6.2.1.4.2, 6.2.4.1 Containment 03/2010 1.12 Nuclear Power Plant Instrumentation for Rev. 2 The APR1400 conforms with this NRC RG. 3.7.4.1 Earthquakes 03/1997 1.13 Spent Fuel Storage Facility Design Basis Rev. 2 The APR1400 conforms with this NRC RG. 9.1.1.1, 9.1.1.3, 03/2007 9.1.2.1, 9.1.3.3.3, 9.1.4.3, 9.1.5.2.1, 9.1.5.3, 9.4.2.1 1.14 Reactor Coolant Pump Flywheel Integrity Rev. 1 The APR1400 conforms with this NRC RG with the 5.4.1.1 08/1975 following exception:

The flywheel hub has small oil channels which prevents accurate ultrasonic test (UT) inspection.

Due to this reason, the ultrasonic inspection for the hub is performed before final machining. Post final machining and during service, the hub is inspected by dye penetrant test or magnetic particle test. Since the flywheel was analyzed using the maximum tensile stress at the outer wheel inner diameter surface, the hub is bounded. The outer wheel will be inspected after machining and during service as indicated in RG 1.14.

1.20 Comprehensive Vibration Assessment Program Rev. 3 The APR1400 conforms with this NRC RG with the 3.9.2.4, for Reactor Internals During Preoperational and 03/2007 following exception: 14.2.7.1.6 Initial Startup Testing

• Startup testing with measurement of SG internals 1.9-7 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (3 of 35)

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.21 Measuring, Evaluating, and Reporting Rev. 2 The APR1400 conforms with this NRC RG. 11.5, Radioactive Material in Liquid and Gaseous 06/2009 12.3.4, Effluents and Solid Waste TS Part 3, 5.0 1.22 Periodic Testing of Protection System Actuation 02/1972 The APR1400 conforms with this NRC RG. 7.1.2.38, Functions Table 7.1-1, 7.2.2.5, 7.2.3.3, 7.3.2.5, 7.3.3.5, 8.1.3.3 1.23 Meteorological Monitoring Programs for Rev. 1 Not applicable (COL) N/A Nuclear Power Plants 03/2007 1.24 Assumptions Used for Evaluating the Potential 03/1972 Not applicable N/A Radiological Consequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure 1.25 Assumptions Used for Evaluating the Potential 03/1972 The APR1400 applies NRC RG 1.183 instead of this N/A Radiological Consequences of a Fuel Handling NRC RG.

Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors 1.26 Quality Group Classifications and Standards for Rev. 4 The APR1400 conforms with this NRC RG. 3.2.2, 3.2.3, 5.2.1.1, Water-, Steam-, and Radioactive-Waste- 03/2007 6.2.4.1.2, 6.6.1, Containing Components of Nuclear Power Plants 9.2.1.2.2.1, 9.5.4.1, 9.5.5.1, 9.5.5.3, 9.5.6, 9.5.6.3, 9.5.6, 9.5.6.3, 9.5.7, 9.5.7.3, 9.5.8, 10.3.6.2, 10.4.2.1, 10.4.3 1.9-8 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (4 of 35)

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.27 Ultimate Heat Sink for Nuclear Power Plants Rev. 2 The APR1400 conforms with this NRC RG with the 9.2.1.1.1, 9.2.5, 01/1976 following exception: 9.2.5.1, 9.2.5.3

• Design of the UHS is site-specific and will be the responsibility of the COL applicant.

1.28 Quality Assurance Program Requirements Rev. 4 The APR1400 conforms with this NRC RG. 10.3.6.2, (Design and Construction) 06/2010 10.4.2.1, 10.4.9.1.2, 14.2.6, 17.5 1.29 Seismic Design Classification Rev. 4 The APR1400 conforms with this NRC RG. 3.2.1, 5.2.5, 03/2007 6.2.4.1.2, 9.1.2.1, 9.1.2.2.3, 9.1.4.3, 9.1.5.2.1, 9.1.5.2.2, 9.1.5.2.3, 9.2.1.1.1, 9.2.2.1.1, 9.2.5.1, 9.4.3.1, 9.4.5.1.2, 9.4.5.1.1, 9.5.4.1, 9.5.5.1, 9.5.5.3, 9.5.6.1, 9.5.6.3, 9.5.7.1, 9.5.7.3, 9.5.8.1, 10.3.1, 10.4.8,10.4.9.3, 11.2, 11.3, 11.4 1.30 Quality Assurance Requirements for the 08/1972 The APR1400 conforms with this NRC RG. 17.5 Installation, Inspection, and Testing of Instrumentation and Electrical Equipment 1.31 Control of Ferrite Content in Stainless Steel Rev. 4 The APR1400 conforms with this NRC RG 4.5.2.2, 4.5.2.4, Weld Metal 10/2013 5.2.3.4.4, 5.3.1.4, 5.4.2.1.4, 6.1.1.1, 6.1.1.2.2, 10.3.6.2 1.9-9 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (5 of 35)

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.32 Criteria for Power Systems for Nuclear Power Rev. 3 The APR1400 conforms with this NRC RG. 8.1.3.3, 8.2.2.2, Plants 03/2004 8.3.1.2.2, 8.3.2.2.2, 9.5.4.1 1.33 Quality Assurance Program Requirements Rev. 3 Not applicable (COL) N/A (Operation) 06/2013 1.34 Control of Electroslag Weld Properties Rev. 1 The APR1400 conforms with this NRC RG except that 5.2.3.3, 5.2.3.4.4, 03/2011 the electroslag process is not used during fabrication of 5.3.1.4, 5.4.2.1.4 any reactor coolant pressure boundary components.

1.35.1 Determining Prestressing Forces for Inspection 07/1990 The APR1400 conforms with this NRC RG. 3.8.1.2.2, of Prestressed Concrete Containments 3.8.1.5.1.2, 3.8.1.5.2.2, 3.8.1.7.2.2, 3.8A.1.4.1.3.3, 3.8.1.2.2, 3.8.1.5.1.2, 3.8.1.5.2.2 1.36 Nonmetallic Thermal Insulation for Austenitic 02/1973 The APR1400 conforms with this NRC RG. 5.2.3.2.3, 5.4.2.1.4, Stainless Steel 6.1.1.1, 6.1.1.2.2, 6.1.1.2.3 1.40 Qualification of Continuous-Duty Safety- Rev. 1 The APR1400 conforms with this NRC RG. N/A Related Motors for Nuclear Power Plants 02/2010 1.41 Preoperational Testing of Redundant On-site 03/1973 The APR1400 conforms with this NRC RG. 14.2.12, Electric Power Systems to Verify Proper Load 8.1.3.3 Group Assignments 1.43 Control of Stainless Steel Weld Cladding of Rev. 1 The APR1400 conforms with this NRC RG. 5.2.3.3, Low-Alloy Steel Components 03/2011 5.3.1.4, 5.4.2.1.3 1.9-10 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (6 of 35)

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.44 Control of the Processing and Use of Stainless Rev. 1 The APR1400 conforms with this NRC RG. 4.5.1.2, 4.5.2.4, Steel 03/2011 5.2.3.2.2, 5.2.3.4.1, 5.3.1.4, 5.4.2.1.4, 6.1.1.1, 6.1.1.2.2, 10.3.6.2 1.45 Guidance on Monitoring and Responding to Rev. 1 The APR1400 conforms with this NRC RG. 3.6.3.5.1, 5.2.5, Reactor Coolant System Leakage 05/2008 5.2.5.1.1.2, 5.2.5.1.2.2, 5.2.5.1.2.4, 5.2.5.1.3, 5.2.5.2, 5.2.5.5, 9.3.3.1.2, 11.5.1.2 1.47 Bypassed and Inoperable Status Indication for Rev. 1 The APR1400 conforms with this NRC RG. 7.1.2.39, Nuclear Power Plant Safety Systems 02/2010 Table 7.1-1, 7.2.2.5, 7.3.2.5, 7.5.1.3, 7.5.2.3, 7.7.1.3, 8.3.1.2.2, 8.1.3.3, 8.3.2.2.2 1.50 Control of Preheat Temperature for Welding of Rev. 1 The APR1400 conforms with this NRC RG. 5.2.3.3, 5.3.1.2, Low-Alloy Steel 03/2011 5.3.1.4, 5.4.2.1.3, 6.1.1.1, 6.1.1.2.2, 10.3.6.2 1.52 Design, Inspection, and Testing Criteria for Air Rev. 4 The APR1400 conforms with this NRC RG. 6.4.2.2, 6.4.6, Filtration and Adsorption Units of Post-Accident 09/2012 6.5.1.1, 6.5.1.1, Engineered-Safety-Feature Atmosphere Cleanup 6.5.1.2.1, 6.5.1.4.1, Systems in Light-Water-Cooled Nuclear Power 6.5.1.4.2, 6.5.1.5, Plants 6.5.1.5.4, 6.5.1.6, 7.3.1.9, 9.4.1.1, 9.4.1.4, 9.4.2.1, 9.4.5.1.3, 9.4.5.4.3, 11.3.4, 14.2.12.1.98 1.9-11 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (7 of 35)

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.53 Application of the Single-Failure Criterion to Rev. 2 The APR1400 conforms with this NRC RG. 7.1.2.40, Nuclear Power Plant Protection Systems 11/2003 Table 7.1-1, 8.1.3.3, 8.3.1.2.2, 8.3.2.2.2 1.54 Service Level I, II, and III Protective Coatings Rev. 2 The APR1400 conforms with this NRC RG. 6.1.2, 11.2.1.2, Applied to Nuclear Power Plants 10/2010 6.8.4.5 1.57 Design Limits and Loading Combinations for Rev. 2 The APR1400 conforms with this NRC RG. 3.8.2.2.2 Metal Primary Reactor Containment System 05/2013 Components 1.59 Design Basis Floods for Nuclear Power Plants Rev. 2 The APR1400 conforms with this NRC RG except for the 2.4, 3.4.1.1 08/1977 actual site-related flooding assessment (COL).

1.60 Design Response Spectra for Seismic Design of Rev. 2 The APR1400 conforms with this NRC RG. 3.7.1.1.1, 3.7.1.1.2, Nuclear Power Plants 07/2014 3.7B-1 1.61 Damping Values for Seismic Design of Nuclear Rev. 1 The APR1400 conforms with this NRC RG. 3.7.1.2, 3.7.3.9, Power Plants 03/2007 3.9.2.2.13, 3.9.3.3.2.2, 3.10.2, 3.10.2.1, 3.10.2.2, 3.12.3.2.1, 3.12.3.2.3, 3.12.3.3, 3.12.5.4, 3.12.6.8, App. 3.9B.4 1.62 Manual Initiation of Protective Actions Rev. 1 The APR1400 conforms with this NRC RG. 7.1.2.41, 06/2010 Table 7.1-1, 10.4.9.1.2, 8.1.3.3 1.63 Electric Penetration Assemblies in Containment Rev. 3 The APR1400 conforms with this NRC RG. 3.11.2, 8.1.3.3, Structures for Nuclear Power Plants 02/1987 Table 8.1-2, 8.3.1.1.9, 8.3.1.2.1, 8.3.1.2.2, 8.3.2.2.1, 8.3.2.2.2 1.9-12 Rev. 2

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.65 Materials and Inspections for Reactor Vessel Rev. 1 The APR1400 conforms with this NRC RG. 5.2.3.6, 5.3.1.7, Closure Studs 04/2010 5.3.3.8 1.68 Initial Test Programs for Water-Cooled Nuclear Rev. 4 The APR1400 conforms with this NRC RG. 6.3.4.2, 7.4.2, Power Plants 06/2013 10.4.4.4, 14.2.7, 14.2.7.1, Table 14.2-7, 14.3.2.14 1.68.2 Initial Startup Test Program to Demonstrate Rev. 2 The APR1400 conforms with this NRC RG. 7.4.2, 14.2.12.4.7 Remote Shutdown Capability for Water-Cooled 04/2010 Nuclear Power Plants 1.68.3 Preoperational Testing of Instrument and Rev. 1 The APR1400 conforms with this NRC RG. 14.2.7.2, Control Air Systems 09/2012 14.2.12.1.125 1.69 Concrete Radiation Shields and Generic Shield Rev. 1 The APR1400 conforms with this NRC RG. 12.3.2.2 Testing for Nuclear Power Plants 05/2009 1.70 Standard Format and Contents of Safety Rev. 3 Not applicable (refer to NRC RG 1.206) N/A Analysis Reports for Nuclear Power Plants 11/1978 1.71 Welder Qualification for Areas of Limited Rev. 1 The APR1400 conforms with this NRC RG. 4.5.2.4, 5.2.3.3, Accessibility 03/2007 5.2.3.4.4, 5.3.1.4, 5.4.2.1.3, 5.4.2.1.4, 6.1.1.1, 6.1.1.2.2, 10.3.6.2 1.72 Spray Pond Piping Made from Fiberglass- Rev. 2 Not applicable N/A Reinforced Thermosetting Resin 11/1978 The APR1400 design does not use spray pond piping made from fiberglass-reinforced thermosetting resin.

1.73 Qualification Tests for Safety-Related Rev. 1 The APR1400 conforms with this NRC RG. 3.9.3.3.1.3, Actuators in Nuclear Power Plants 10/2013 3.9.3.3.1.3.1, 3.11.2, 8.1.3.3 1.9-13 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (9 of 35)

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.75 Criteria for Independence of Electrical Safety Rev. 3 The APR1400 conforms with this NRC RG except the 7.1.2.42, Systems 02/2005 following. Table 7.1-1,

• Two CEA position inputs instead of four CEA 7.2.2.3, 7.3.2.3, position input described in Subsection 7.1.2.3. 7.9.2.7, 8.1.3.3, 8.3.1.1.2.3, 8.3.1.1.9, 8.3.1.1.10, 8.3.1.2.2, 8.3.2.1.2.2, 8.3.2.1.2.4, 8.3.2.1.2.5, 8.3.2.2.2 1.76 Design-Basis Tornado and Tornado Missiles Rev. 1 The APR1400 conforms with this NRC RG. Table 2.0-1, for Nuclear Power Plants 03/2007 3.3.2.1, 3.5.1.4 1.77 Assumptions Used for Evaluating a Control 05/1974 The APR1400 conforms with this NRC RG. Refer to 15.4.8 Rod Ejection Accident for Pressurized Water Subsection 15.4.8 for further information.

Reactors Note: SRP Section 4.2 Appendix B will be used in conjunction with the requirements of NRC RG 1.77 for the APR1400.

1.78 Evaluating the Habitability of a Nuclear Power Rev. 1 The APR1400 conforms with this NRC RG except for the 6.4.4.2, 6.4.7 Plant Control Room During a Postulated 12/2001 following:

Hazardous Chemical Release

• Full conformance by the COL applicant with site-specific consequence data.

1.79 Preoperational Testing of Emergency Core Rev. 2 The APR1400 conforms with this NRC RG. 6.3.4.1, 14.2.7.3 Cooling Systems for Pressurized Water 10/2013 Systems 1.81 Shared Emergency and Shutdown Electric Rev. 1 Not applicable. N/A Systems for Multi-Unit Nuclear Power Plants 01/1975 The APR1400 is a single unit plant; therefore, this NRC RG is not applicable to the APR1400.

1.9-14 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (10 of 35)

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.82 Water Sources for Long-Term Recirculation Rev. 4 The APR1400 conforms with this NRC RG. 6.2.2.2.5, 6.2.2.3, Cooling Following a Loss-of-Coolant Accident 03/2012 6.3.1.3, 6.3.2.2.3, 6.8.2.2.1, 6.8.4.5, Table 15.0-12, Table 15.0-13 1.84 Design, Fabrication, and Materials Code Case Rev. 36 The APR1400 conforms with this NRC RG. 3.12.2.2, 3.13.1.1, Acceptability, ASME Section III 08/2014 4.5.1.1, 4.5.2.1, 5.2.3.1, 5.2.6, 6.0, 10.3.6.2 1.86 Termination of Operating Licenses for Nuclear 06/1974 Not applicable (COL) N/A Reactors 1.87 Guidance for Construction of Class 1 06/1975 Not applicable N/A Components in Elevated Temperature Reactors 1.89 Environmental Qualification of Certain Electric Rev. 1 The APR1400 conforms with this NRC RG except for the 3.9.3.3.1.2, Equipment Important to Safety for Nuclear 06/1984 following. 3.9.3.3.1.3.1, Power Plants

• IEEE Standard 323-2003 is applied instead of 3.11.2, 3.11.5.2, IEEE Standard 323-1974 because NRC RG 1.209 Table 6.5-2, endorses the current national qualification Table 7.1-1 standard (IEEE Standard 323-2003).

1.90 Inservice Inspection of Prestressed Concrete Rev. 2 Not applicable N/A Containment Structures with Grouted Tendons 11/2012 The APR1400 adopted prestressed concrete containment with ungrouted tendon.

1.91 Evaluations of Explosions Postulated to Occur Rev. 2 Not applicable (COL) N/A on Transportation Routes Near Nuclear Power 04/2013 Plants 1.9-15 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (11 of 35)

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.92 Combining Modal Responses and Spatial Rev. 3 The APR1400 conforms with this NRC RG. 3.7.2.6, 3.7.2.7, Components in Seismic Response Analysis 10/2012 3.7.3.5, 3.9.2.2.5, 3.9.2.2.6, 3.12.3.2.4, 3.12.3.2.5, 3.12.5.5, App. 3.9B.4 1.93 Availability of Electric Power Sources Rev. 1 The APR1400 conforms with this NRC RG. 8.1.3.3, 16.3.8 03/2012 1.96 Design of Main Steam Isolation Valve Leakage Rev. 1 Not applicable (BWR) N/A Control Systems for Boiling Water Reactor 06/1976 Nuclear Power Plants 1.97 Criteria for Accident Monitoring Instrumentation Rev. 4 The APR1400 conforms with this NRC RG. 1.2.6.3, 3.10, for Nuclear Power Plants. 06/2006 3.11.1.1, 3.11.2, 3.11.6, 7.1.1.5, 7.1.2.43, Table 7.1-1, 7.5.1.1, 7.5.2.1, 7.7.1.2, 10.4.9.5.3, 10.4.9.5, 11.5.1, 11.5.2, 11.5.2.1, 11.5.2.2.3.1, 12.3.1.7, 12.3.4.1.1, 12.3.4.1.5, 14.3.2.7 1.98 Assumptions Used for Evaluating the Potential 03/1976 Not applicable (BWR) N/A Radiological Consequences of a Radioactive Offgas System Failure in a Boiling Water Reactor 1.9-16 Rev. 2

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.99 Radiation Embrittlement of Reactor Vessel Rev. 2 The APR1400 conforms with this NRC RG. 5.2.3.1, 5.3.1.4, Materials 05/1988 5.3.1.6.7, 5.3.2, 5.3.2.1.1, 5.3.2.1.2, 5.3.2.4 1.100 Seismic Qualification of Electric and Mechanical Rev. 3 The APR1400 conforms with this NRC RG. 3.9.2.2.1, Equipment for Nuclear Power Plants 09/2009 3.9.3.3.1.1, 3.9.3.3.1.2, 3.9.3.3.1.3, 3.9.3.3.1.3.1, 3.9.3.3.2.2, 3.9.6.1, 3.10.1.1, 3.10.1, 3.10.2, 3.10.2.1, 3.10.2.2, 3.10.2.3, 3.11.2, 5.2.2.1.1, 5.4.12.2.1, 5.4.12.2.2, Table 6.5-2, 8.3.2.2.2 1.101 Emergency Planning and Preparedness for Rev. 5 Not applicable (COL) N/A Nuclear Power Reactors 06/2005 1.102 Flood Protection for Nuclear Power Plants Rev. 1 The APR1400 conforms with this NRC RG. 3.4.1.1, 3.4.1.2 09/1976 1.105 Setpoints for Safety-Related Instrumentation Rev. 3 The APR1400 conforms with this NRC RG except for the 7.1.2.44, 12/1999 setpoints calculation using CPC Setpoint Analysis Table 7.1-1, Methodology Technical Report (Reference 8). 7.2.2.7, 7.3.2.7, 15.0.0.9 1.106 Thermal Overload Protection for Electric Motors Rev. 2 The APR1400 conforms with this NRC RG. Table 8.1-2, on Motor-Operated Valves 02/2012 8.3.1.1.3.11, 8.3.1.2.2, 8.3.2.2.2 1.9-17 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (13 of 35)

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.107 Qualifications for Cement Grouting for Rev. 2 Not applicable N/A Prestressing Tendons in Containment Structures 06/2011 The APR1400 adopted prestressed concrete containment with ungrouted tendon.

1.109 Calculation of Annual Doses to Man From Rev. 1 The APR1400 conforms with this NRC RG. 11.2.5, Routine Releases of Reactor Effluents for the 10/1977 Table 11.2-4, Purpose of Evaluating Compliance with 10 CFR 11.3.3.1, 11.3.7, Part 50, Appendix I Table 11.3-5, 11.4.4, 11.5.1.2 1.110 Cost-Benefit Analysis for Radwaste Systems for Rev. 1 The APR1400 conforms with this NRC RG. 11.2.1.5, 11.3.1.6, Light-Water-Cooled Nuclear Power Reactors 10/2013 The cost-benefit analysis approach stipulated by 10 CFR Part 50, Appendix I, Section II, Paragraph D requires that a population dose analysis be performed to demonstrate that the radwaste system is designed consistent with the as low as reasonably achievable criterion.

Due to the extreme site-specific nature of population dose analyses, the cost-benefit analysis is deferred to site-specific environmental reports.

1.111 Methods for Estimating Atmospheric Transport Rev. 1 Not applicable (COL) N/A and Dispersion of Gaseous Effluents in Routine 07/1977 Releases From Light-Water-Cooled Reactors 1.112 Calculation of Releases of Radioactive Materials Rev. 1 The APR1400 conforms with this NRC RG. 11.1, 11.2.2, in Gaseous and Liquid Effluents from Light- 03/2007 11.2.3.1, 11.3.1.2 Water-Cooled Power Reactors 1.113 Estimating Aquatic Dispersion of Effluents from Rev. 1 The APR1400 conforms with this NRC RG. 11.2.5, 11.4.4, Accidental and Routine Reactor Releases for the 04/1977 11.5.1.2 Purpose of Implementing Appendix I 1.9-18 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (14 of 35)

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.114 Guidance to Operators at the Controls and to Rev. 3 Not applicable (COL) N/A Senior Operators in the Control Room of a 10/2008 Nuclear Power Plant 1.115 Protection Against Turbine Missiles Rev. 2 The APR1400 conforms with this NRC RG. 3.5.1.3, 9.1.2, 01/2012 10.1.1, 10.2.1, 10.3.1 1.117 Tornado Design Classification Rev. 1 The APR1400 conforms with this NRC RG. 3.3.2, 9.1.2.1, 04/1978 10.3.1 1.118 Periodic Testing of Electric Power and Rev. 3 The APR1400 conforms with this NRC RG. Table 6.5-2, Protection Systems 04/1995 7.1.2.45, Table 7.1-1, 7.2.3.3, 7.3.2.5, 7.5.2.1, 7.6.2.3, Table 8.1-2, 8.1.3.3, 8.3.1.1.6, 8.3.1.2.2, 8.3.2.2.2 1.121 Bases for Plugging Degraded PWR Steam 08/1976 The APR1400 conforms with this NRC RG. 5.4.2, 5.4.2.2.2.12 Generator Tubes 1.122 Development of Floor Design Response Spectra Rev. 1 The APR1400 conforms with this NRC RG. 3.7.2.5, for Seismic Design of Floor-Supported 02/1978 3.7.2.9, 3.7A.3.3, Equipment or Components 3.12.3.2.2 1.124 Service Limits and Loading Combinations for Rev. 3 The APR1400 conforms with this NRC RG. 3.9.3.4 Class 1 Linear Type Component Supports 07/2013 Table 3.9-3 1.125 Physical Models for Design and Operation of Rev. 2 Not applicable (COL) N/A Hydraulic Structures and Systems for Nuclear 03/2009 Power Plants 1.9-19 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (15 of 35)

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.126 An Acceptable Model and Related Statistical Rev. 2 The APR1400 conforms with this NRC RG. 4.2.1.2.3 Methods for the Analysis of Fuel Densification 03/2010 1.127 Inspection of Water-Control Structures Rev. 1 Not applicable (COL) N/A Associated with Nuclear Power Plants 03/1978 1.128 Installation Design and Installation of Vented Rev. 2 The APR1400 conforms with this NRC RG. 8.1.3.3, Lead-Acid Storage Batteries for Nuclear Power 02/2007 Table 8.1-2, Plants 8.3.2.2.2, 9.4.4.2.2, 9.4.5.2.2.3, 9.4.5.2.2.4, 9.4.7.2.1 1.129 Maintenance, Testing, and Replacement of Rev. 3 The APR1400 conforms with this NRC RG. 8.1.3.3, Vented Lead-Acid Storage Batteries for Nuclear 09/2013 Table 8.1-2, Power Plants 8.3.2.2.2 1.130 Service Limits and Loading Combinations for Rev. 3 The APR1400 conforms with this NRC RG. 3.9.3.4 Class 1 Plate-and-Shell-Type Supports 07/2013 Table 3.9-3 1.132 Site Investigations for Foundations of Nuclear Rev. 2 Not applicable (COL) N/A Power Plants 10/2003 1.133 Loose-Part Detection Program for the Primary Rev. 1 The APR1400 conforms with this NRC RG. 7.7.1.5 Systems of Light-Water-Cooled Reactors 05/1981 1.134 Medical Evaluation of Licensed Personnel at Rev. 3 Not applicable (COL) N/A Nuclear Power Plants 03/1998 1.136 Design Limits, Loading Combinations, Rev. 3 The APR1400 conforms with this NRC RG. 3.8.1.2.2, 3.8.1.3, Materials, Construction, and Testing of Concrete 03/2007 3.8.1.3.2, 3.8.1.4.7, Containments 3.8.1.6, 3.8.1.6.3, 3.8A.1.3.1, 3.8A.1.3.2 1.9-20 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (16 of 35)

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.137 Fuel Oil Systems for Emergency Power Supplies Rev. 2 The APR1400 conforms with this NRC RG. 8.1.3.3, 06/2013 9.5.4.1 1.138 Laboratory Investigations of Soils and Rocks for Rev. 2 Not applicable (COL) N/A Engineering Analysis and Design of Nuclear 12/2003 Power Plants 1.140 Design, Inspection, and Testing Criteria for Air Rev. 2 The APR1400 conforms with this NRC RG. 9.4.2.1, Filtration and Adsorption Units of Normal 06/2001 9.4.2.4, 9.4.5.1.3, Atmosphere Cleanup Systems in Light-Water- 9.4.5.4.3, 9.4.6.4.2, Cooled Nuclear Power Plants 9.4.7.4, 11.3.4, 14.2.12.1.94, 14.2.12.1.98, 14.2.12.1.99 1.141 Containment Isolation Provisions for Fluid Rev. 1 The APR1400 conforms with this NRC RG. 6.2.4 Systems 07/2010 1.142 Safety-Related Concrete Structures for Nuclear Rev. 2 The APR1400 conforms with this NRC RG. 3.5.3.2, 3.8.3.3, Power Plants 11/2001 3.8.4.4, 3.8.4.4.2.1, 3.8.4.5, 3.8.5.4, 3.8A.2.4.2 1.143 Design Guidance for Radioactive Waste Rev. 2 The APR1400 conforms with this NRC RG. 3.2.1, 3.2.2, Management Systems, Structures, and 11/2001 10.4.8.1.2, 11.2, Components Installed in Light-Water-Cooled 11.2.1.2, 11.2.2.3, Nuclear Power Plants 11.2.5, 11.3, 11.3.1.3, 11.3.5, Table 11.3-2, 11.4.1.2, 11.4.1.3, 11.4.1.7, 11.4.2.2.1, 11.4.2.3, 11.4.9, 11.5.1.2 1.9-21 Rev. 2

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.145 Atmospheric Dispersion Models for Potential Rev. 1 Not applicable (COL) N/A Accident Consequence Assessment at Nuclear 02/1983 Power Plants 1.147 In-service Inspection Code Case Acceptability, Rev. 17 Not applicable (COL) 5.2.1.2, 5.2.4.1.8, ASME Section XI, Division 1 08/2014 5.2.4.2, 6.6.1, 6.6.3 1.149 Nuclear Power Plant Simulation Facilities for Rev. 4 The APR1400 conforms with this NRC RG. N/A Use in Operator Training, License Examinations, 04/2011 and Applicant Experience Requirements 1.151 Instrument Sensing Lines Rev. 1 The APR1400 conforms with this NRC RG. 3.2.1, 07/2010 7.1.2.46, Table 7.1-1, 7.2.2.3, 7.3.2.3 1.152 Criteria for Digital Computer in Safety Systems Rev. 3 The APR1400 conforms with this NRC RG. 1.5.4, of Nuclear Power Plants 07/2011 7.1.2.47, 7.1.2.71, Table 7.1-1, 7.9.1.2 1.153 Criteria for Safety Systems Rev. 1 The APR1400 conforms with this NRC RG. 8.1.3.3, Table 8.1-2, 06/1996 8.3.1.1.2.2, 8.3.1.2.2, 8.3.2.1.2.4, 8.3.2.2.2 1.155 Station Blackout 08/1988 The APR1400 conforms with this NRC RG. 8.1.2, 8.1.3.2, 8.1.3.3 Table 8.1-2, 8.2.2.2, 8.3.1.1.3, 8.3.1.2.2, 8.3.2.2.2, 8.4.1.1, 8.4.1.2, 8.4.1.3, 8.4.1.6, 8.4.2.2, 9.4.1.1, 9.5.9, 9.5.9.1, 9.5.9.2, 10.3.1, 10.4.9.1.2 1.9-22 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (18 of 35)

Revision / DCD Tier 2 NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation Section 1.156 Qualification of Connection Assemblies for Rev. 1 The APR1400 conforms with this NRC RG. 3.11.2, 8.1.3.3 Nuclear Power Plants 07/2011 1.157 Best-Estimate Calculations of Emergency Core 05/1989 The APR1400 conforms with this NRC RG. SBLOCA 6.2.1.5.1, Cooling System Performance analyses performed for the APR1400 are based on the 15.6.5.3.1 more stringent requirements of Appendix K to 10 CFR Part 50.

1.158 Qualification of Safety-Related Lead Storage 02/1989 The APR1400 conforms with this NRC RG except for the 3.11.2, 8.1.3.3 Batteries for Nuclear Power Plants following.

• IEEE Standard 535-2013 is applied instead of IEEE Standard 535-1986 because IEEE Std. 535-1986 does not apply to duty cycles longer than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

1.159 Assuring the Availability of Funds for Rev. 2 Not applicable (COL) N/A Decommissioning Nuclear Reactors 10/2011 1.160 Monitoring the Effectiveness of Maintenance at Rev. 3 Not applicable (COL) N/A Nuclear Power Plants 05/2012 1.161 Evaluation of Reactor Pressure Vessels with 06/1995 Not applicable N/A Charpy Upper Shelf Energy Less Than 50 Ft-Lb 1.162 Format and Contents of Report for Thermal 02/1996 Not applicable (COL) N/A Annealing of Reactor Pressure Vessel 1.163 Performance-Based Containment Leak-Test 09/1995 The APR1400 conforms with this NRC RG. 6.2.1.6, 6.2.6, Program 6.2.6.1, 6.2.6.4 1.166 Pre-Earthquake Planning and Immediate Nuclear 03/1997 Not applicable (COL) N/A Power Plant Operator Postearthquake Action 1.9-23 Rev. 2

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.167 Restart of Nuclear Power Plant Shut Down by a 03/1997 Not applicable (COL) N/A Seismic Event 1.168 Verification, Validation, Reviews, and Audits for Rev. 2 The APR1400 conforms with this NRC RG. 7.1.2.48, Digital Computer Software Used in Safety 07/2013 Table 7.1-1 System of Nuclear Power Plants 1.169 Configuration Management Plans for Digital Rev. 1 The APR1400 conforms with this NRC RG except the 7.1.2.49, Computer Software Used in Safety Systems of 07/2013 position as follows; Table 7.1-1 Nuclear Power Plants

• Regulatory Position

- Conformance with IEEE 828-2005

• Position of the APR1400

- Conformance with IEEE 828-1998

• Summary Description of Deviation

- There are no deviations in the body between IEEE 828-1998 and IEEE 828-2005 except for the document numbering and adding Appendix B in IEEE 828-1998. Therefore, conformance with IEEE 828-2005 is met by conforming with IEEE 828-1998.

1.170 Software Test Documentation for Digital Rev. 1 The APR1400 conforms with this NRC RG. 7.1.2.50, Computer Software Used in Safety Systems of 07/2013 Table 7.1-1 Nuclear Power Plants 1.171 Software Unit Testing for Digital Computer Rev. 1 The APR1400 conforms with this NRC RG. 7.1.2.51, Software Used in Safety Systems of Nuclear 07/2013 Table 7.1-1 Power Plants 1.172 Software Requirements Specifications for Digital Rev. 1 The APR1400 conforms with this NRC RG. 7.1.2.52, Computer Software Used in Safety Systems of 07/2013 Table 7.1-1 Nuclear Power Plants 1.9-24 Rev. 2

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.173 Developing Software Life Cycle Processes for Rev. 1 The APR1400 conforms with this NRC RG. 7.1.2.53, Digital Computer Software Used in Safety 07/2013 Table 7.1-1 Systems of Nuclear Power Plants 1.174 An Approach for Using Probabilistic Risk Rev. 2 Not applicable N/A Assessment in Risk-Informed Decisions on 05/2011 Plant-Specific Changes to the Licensing Basis 1.175 An Approach for Plant-Specific, Risk-Informed 08/1998 Not applicable N/A Decisionmaking: Inservice Testing 1.177 An Approach for Plant-Specific, Risk-Informed Rev. 1 Not applicable N/A Decisionmaking: Technical Specifications 05/2011 1.178 An Approach for Plant-Specific Risk-Informed Rev. 1 Not applicable N/A Decisionmaking for Inservice Inspection of 09/2003 Piping 1.179 Standard Format and Content of License Rev. 1 Not applicable (COL) N/A Termination Plans for Nuclear Power Reactors 06/2011 1.180 Guidelines for Evaluating Electromagnetic and Rev. 1 The APR1400 conforms with this NRC RG. 3.11.2, Radio-Frequency Interference in Safety-Related 10/2003 7.1.2.54, Instrumentation and Control Systems Table 7.1-1, 7.2.2.8, 7.3.2.8, 7.9.2.11, 8.1.3.3 1.181 Content of the Updated Final Safety Analysis 09/1999 Not applicable (COL) N/A Report in Accordance with 10 CFR 50.71(e) 1.9-25 Rev. 2

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.183 Alternative Radiological Source Terms for 07/2000 The APR1400 conforms with this NRC RG. 3.11.2, 3.11.5.2, Evaluating Design Basis Accidents at Nuclear 6.4.2.5, Table 6.5-2, Power Reactors 12.2.3, 12.3.1.8, 12.4.1.2.7, 15.0.3.2, 15.0.3.3, 15.0.3.4, 15.0.3.6, 15.0.3.7, 15.1.5.5, 15.1.5.5.2, 15.2.8.5.2, 15.3.3.5.1, 15.3.3.5.2, 15.4.8.5.1, 15.4.8.5.2, 15.6.2.5, 15.6.3.2.5, 15.6.3.2.5.2, 15.6.5.5, 15.6.5.5.1.1, 15.6.5.5.1.2, 15.6.5.5.1.3, 15.7.4.1, 15.7.4.2, 15A.1.2.2, 15A.2.1, 15A.5.2.1, 15A.5.2.2, 15A.5.3 1.184 Decommissioning of Nuclear Power Reactors Rev. 1 Not applicable (COL) N/A 10/2013 1.185 Standard Format and Content for Post-Shutdown Rev. 1 Not applicable (COL) N/A Decommissioning Activities Report 06/2013 1.186 Guidance and Examples for Identifying 10 CFR 12/2000 Not applicable N/A 50.2 Design Bases 1.187 Guidance for Implementation of 10 CFR 50.59, 11/2000 Not applicable (COL) N/A Changes, Tests, and Experiments 1.188 Standard Format and Content for Applications To Rev. 1 Not applicable (COL) N/A Renew Nuclear Power Plant Operating Licenses 09/2005 1.9-26 Rev. 2

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.189 Fire Protection for Nuclear Power Plants Rev. 2 Conformance with exceptions. 3.2.1, 10/2009 Refer to Table 9.5.1-1 for a point-by-point discussion of 7.1.2.55, Table 7.1-conformance with this guide. 1, 7.4.2, 8.1.3.3, 9.5.1.1, 9.5.1.2, 9.5.1.2.1, 9.5.1.2.1, 9.5.1.2.3, 9.5.1.2.4, 9.5.1.2.5, 9.5.1.2.6, 9.5.1.3.2, 9.5.2.2.1.8, 9.5.3.1, Table 9.5.1-1, Table 9.5.1-2, 9.5A.1, 9.5A.2.1, 9.5A.2.5, 9.5A.2.5.1, 9.5A.3.1, 11.3.1.4 1.190 Calculational and Dosimetry Methods for 03/2001 The APR1400 conforms with this NRC RG. N/A Determining Pressure Vessel Neutron Fluence 1.191 Fire Protection Program for Nuclear Power Plants 05/2001 Not applicable (COL) N/A During Decommissioning and Permanent Shutdown 1.192 Operation and Maintenance Code Case Rev. 1 The APR1400 conforms with this NRC RG. 5.2.1.2 Acceptability, ASME OM Code 08/2014 1.193 ASME Code Cases Not Approved for Use Rev. 4 The APR1400 conforms with this NRC RG. N/A 08/2014 1.194 Atmospheric Relative Concentrations for Control 06/2003 The APR1400 conforms with this NRC RG. 2.3.4 Room Radiological Habitability Assessments at Nuclear Power Plants 1.195 Methods and Assumptions for Evaluating 05/2003 The APR1400 applies to NRC RG 1.183 instead of this N/A Radiological Consequences of Design Basis NRC RG.

Accidents at Light-Water Nuclear Power Reactors 1.9-27 Rev. 2

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.196 Control Room Habitability at Light-Water Rev. 1 The APR1400 conforms with this NRC RG. 6.4 Nuclear Power Reactors 01/2007 1.197 Demonstrating Control Room Envelope Integrity 05/2003 The APR1400 conforms with this NRC RG. 6.4.5, 9.4.1.4 at Nuclear Power Reactors 1.198 Procedures and Criteria for Assessing Seismic 11/2003 Not applicable (COL) N/A Soil Liquefaction at Nuclear Power Plant Sites 1.199 Anchoring Components and Structural Supports 11/2003 The APR1400 conforms with this NRC RG. 3.8.4, 4.2.1, in Concrete 3.8.4.4.2.6, 3.12.6.4 1.200 An Approach for Determining the Technical Rev. 2 The APR1400 conforms with this NRC RG. Table 7.1-1, Adequacy of Probabilistic Risk Assessment 03/2009 19.0, 19.1 Results for Risk-Informed Activities 1.201 Guidelines for Categorizing Structures, Systems, Rev. 1 Not applicable. N/A and Components in Nuclear Power Plants 05/2006 This NRC RG was written to address PRAs performed in According to Their Safety Significance support of changes proposed for existing, already-licensed plants.

1.202 Standard Format and Content of 02/2005 Not applicable (COL) N/A Decommissioning Cost Estimates for Nuclear Power Reactors 1.203 Transient and Accident Analysis Methods 12/2005 Not applicable. N/A APR1400 conforms with the NRC RG 1.157.

1.204 Guidelines for Lightning Protection of Nuclear 11/2005 The APR1400 conforms with this NRC RG. 7.1.2.56, Power Plants Table 7.1-1, Table 8.1-2, 8.1.3.3, 8.2.2.2, 8.3.1.1.8, 8.3.1.2.2.

15.6.5 1.9-28 Rev. 2

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NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.205 Risk-Informed, Performance-Based Fire Rev. 1 Not applicable. N/A Protection for Existing Light-Water Nuclear 12/2009 This NRC RG is allowed to the nuclear power reactor Power Plants licensees to permit reactor licensees to adopt risk-informed, performance-based approach as an alternative to the existing deterministic fire protection requirement.

APR 1400 fire protection is designed to the requirements of NRC RG 1.189, Rev. 2, which provides deterministic fire protection guidance; it is not necessary to incorporate this NRC RG.

1.206 Combined License Applications for Nuclear 06/2007 The APR1400 conforms with exception. All Power Plants (LWR Edition) Section C.II.2, C.III, and C.IV are guidance for COL application referencing a certified design and/or an early site permit (ESP). These sections conform in the COL application.

The APR1400 is not a passive-ALWR-plant. Therefore, Section C.IV.9 is not applicable to the APR1400.

1.207 Guidelines for Evaluating Fatigue Analyses 03/2007 The APR1400 conforms with this NRC RG. 3.9.1.1, Incorporating the Life Reduction of Metal 3.9.1.2.1.12, Components Due to the Effects of the Light- 3.9.3.1, 3.12.5.7, Water Reactor Environment for New Reactors 3.12.5.19 1.208 A Performance-Based Approach to Define the 03/2007 Not applicable (COL) N/A Site-Specific Earthquake Ground Motion 1.209 Guidelines for Environmental Qualification of 03/2007 The APR1400 conforms with this NRC RG. 7.2.2.8 Safety-Related Computer-Based Instrumentation and Control Systems in Nuclear Power Plants 1.210 Qualification of Safety-Related Battery Chargers 06/2008 The APR1400 conforms with this NRC RG. 3.11, 8.1.3.3 and Inverters for Nuclear Power Plants 1.9-29 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (25 of 35)

Revision /

NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.211 Qualification of Safety-Related Cables and Field 04/2009 The APR1400 conforms with this NRC RG. 3.11.2, 8.1.3.3 Splices for Nuclear Power Plants 1.212 Sizing of Large Lead-Acid Storage Batteries 11/2008 The APR1400 conforms with this NRC RG except for the 8.1.3.3, 8.3.2.2.2 following.

• IEEE Standard 485-2010 is applied instead of IEEE Standard 485-1997 because NRC RG 1.129 endorses the current standard (IEEE Standard 485-2010).

1.213 Qualification of Safety-Related Motor Control 05/2009 The APR1400 conforms with this NRC RG. 3.11, 8.1.3.3 Centers for Nuclear Power Plants 1.215 Guidance for ITAAC Closure Under 10 CFR 52 Rev. 1 The APR1400 conforms with this NRC RG. 14.3.2.3, 14.3.5 05/2012 1.216 Containment Structural Integrity Evaluation for 08/2010 The APR1400 conforms with this NRC RG. 19.2.1, Internal Pressure Loadings Above Design-Basis 19.2.4.2.2 Pressure 1.217 Guidance for the Assessment of Beyond-Design- 08/2011 The APR1400 conforms with this NRC RG. 19.5 Basis Aircraft Impacts 1.218 Condition-Monitoring Techniques for Electric 04/2012 Not applicable (COL) N/A Cables Used in Nuclear Power Plants 1.219 Guidance on Making Changes to Emergency 11/2011 Not applicable (COL) N/A Plans for Nuclear Power Reactors 1.221 Design-Basis Hurricane and Hurricane Missiles 10/2011 The APR1400 conforms with this NRC RG. Table 2.0-1, for Nuclear Power Plants 3.3.2.1, 3.5.1.4 4.1 Radiological Environmental Monitoring for Rev. 2 Not applicable (COL) N/A Nuclear Power Plants 06/2009 1.9-30 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (26 of 35)

Revision /

NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 4.2 Preparation of Environmental Reports for Rev. 2 Not applicable (COL) N/A Nuclear Power Stations 07/1976 4.2 Supplement 1 - Preparation of Supplemental 09/2000 Not applicable (COL) N/A Environmental Reports for Applications To Renew Nuclear Power Plant Operating Licenses 4.7 General Site Suitability Criteria for Nuclear Rev. 3 Not applicable (COL) N/A Power Stations 03/2014 4.9 Preparation of Environmental Reports for Rev. 1 Not applicable (COL) N/A Commercial Uranium Enrichment Facilities 10/1975 4.11 Terrestrial Environmental Studies for Nuclear Rev. 2 Not applicable (COL) N/A Power Stations 03/2012 4.13 Performance, Testing, and Procedural Rev. 1 Not applicable (COL) N/A Specifications for Thermo luminescence 07/1977 Dosimetry: Environmental Applications 4.14 Radiological Effluent and Environmental Rev. 1 Not applicable N/A Monitoring at Uranium Mills 04/1980 4.15 Quality Assurance for Radiological Monitoring Rev. 2 Not applicable (COL) N/A Programs (Inception through Normal Operations 07/2007 to License Termination) - Effluent Streams and the Environment 4.16 Monitoring and Reporting Radioactive Materials Rev. 2 Not applicable N/A in Liquid and Gaseous Effluents from Nuclear 12/2010 Fuel Cycle Facilities 4.17 Standard Format and Content of Site Rev. 1 Not applicable N/A Characterization Plans for High-Level-Waste 03/1987 Geologic Repositories 1.9-31 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (27 of 35)

Revision /

NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 4.18 Standard Format and Content of Environmental 06/1983 Not applicable N/A Reports for Near-Surface Disposal of Radioactive Waste 4.19 Guidance for Selecting Sites for Near-Surface 08/1988 Not applicable N/A Disposal of Low-Level Radioactive Waste 4.20 Constraint on Releases of Airborne Radioactive Rev. 1 Not applicable N/A Materials to the Environment for Licensees other 04/2012 than Power Reactors 4.21 Minimization of Contamination and Radioactive 06/2008 The APR1400 conforms with this NRC RG 5.4.3.5, 5.4.7.4.4, Waste Generation: Life-Cycle Planning 5.4.12.2.3, 6.2.5.2.3, 6.3.6, 6.5.2.2.1, Table 6.5-2, 6.8.2.1.3, 9.1.2.2.2, 9.1.3.2.3, 9.2.1.2.4, 9.2.2.2.5, 9.3.2.2.4, 9.3.3.2.6, 9.3.4.2.10, 9.4.6.1.1, 9.4.8, 10.3.2.4, 10.4.2.2.3, 10.4.6.2.4, 10.4.7.2.4, 10.4.8.2.4, 10.4.9.2.5, 11.2.1.1, 11.2.1.2, 11.2.2.4.1, 11.3.1.2, 11.3.2.2.2, 11.4, 11.4.2.5.1, 11.5.2.4, 12.1.3.1, 12.1.3.4, 12.3.1.1, 12.3.1.4, 12.3.1.10, 12.3.1.10, 12.3.1.10.1, 12.3.1.10.2, 12.3.1.10.3, Table 12.3-6, Table 12.3-7 1.9-32 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (28 of 35)

Revision /

NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 4.22 Decommissioning Planning During Operations 12/2012 Not applicable (COL) N/A 5.3 Statistical Terminology and Notation for Special 02/1973 Not applicable N/A Nuclear Materials Control and Accountability 5.4 Standard Analytical Methods for the 02/1973 Not applicable N/A Measurement of Uranium Tetrafluoride (UF4) and Uranium Hexafluoride (UF6) 5.5 Standard Methods for Chemical, Mass 02/1973 Not applicable N/A Spectrometric, and Spectrochemical Analysis of Nuclear-Grade Uranium Dioxide Powders and Pellets 5.7 Entry/Exit Control for Protected Areas, Vital Rev. 1 Not applicable (COL) N/A Areas, and Material Access Areas 05/1980 5.8 Design Considerations for Minimizing Residual Rev. 1 Not applicable N/A Holdup of Special Nuclear Material in Drying 05/1974 and Fluidized Bed Operations 5.9 Guidelines for Germanium Spectroscopy Rev. 2 Not applicable N/A Systems for Measurement of Special Nuclear 12/1983 Material 5.11 Nondestructive Assay of Special Nuclear Rev. 1 Not applicable N/A Material Contained in Scrap and Waste 04/1984 5.12 General Use of Locks in the Protection and 11/1973 Not applicable N/A Control of Facilities and Special Nuclear Materials 5.13 Conduct of Nuclear Material Physical 11/1973 Not applicable N/A Inventories 1.9-33 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (29 of 35)

Revision /

NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 5.18 Limit of Error Concepts and Principles of 01/1974 Not applicable N/A Calculation in Nuclear Materials Control 5.20 Training, Equipping, and Qualifying of Guards 01/1974 Not applicable N/A and Watchmen 5.21 Nondestructive Uranium-235 Enrichment Assay Rev. 1 Not applicable N/A by Gamma Ray Spectrometry 12/1983 5.22 Assessment of the Assumption of Normality 04/1974 Not applicable N/A (Employing Individual Observed Values) 5.23 In Situ Assay of Plutonium Residual Holdup Rev. 1 Not applicable N/A 02/1984 5.25 Design Considerations for Minimizing Residual 06/1974 Not applicable N/A Holdup of Special Nuclear Material in Equipment for Wet Process Operations 5.26 Selection of Material Balance Areas and Item Rev. 1 Not applicable N/A Control Areas 04/1975 5.27 Special Nuclear Material Doorway Monitors 06/1974 Not applicable N/A 5.28 Evaluation of Shipper-Receiver Differences in 06/1974 Not applicable N/A the Transfer of Special Nuclear Materials 5.31 Specially Designed Vehicle with Armed Guards Rev. 1 Not applicable N/A for Road Shipment of Special Nuclear Material 04/1975 5.33 Statistical Evaluation of Material Unaccounted 06/1974 Not applicable N/A For 5.34 Nondestructive Assay for Plutonium in Scrap Rev. 1 Not applicable N/A Material by Spontaneous Fission Detection 05/1984 1.9-34 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (30 of 35)

Revision /

NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 5.36 Recommended Practice for Dealing with 06/1974 Not applicable N/A Outlying Observations 5.37 In Situ Assay of Enriched Uranium Residual Rev. 1 Not applicable N/A Holdup 10/1983 5.38 Nondestructive Assay of High-Enrichment Rev. 1 Not applicable N/A Uranium Fuel Plates by Gamma Ray 10/1983 Spectrometry 5.39 General Methods for the Analysis of Uranyl 12/1974 Not applicable N/A Nitrate Solutions for Assay, Isotopic Distribution, and Impurity Determinations 5.42 Design Considerations for Minimizing Residual 01/1975 Not applicable N/A Holdup of Special Nuclear Material in Equipment for Dry Process Operations 5.43 Plant Security Force Duties 01/1975 Not applicable N/A 5.44 Perimeter Intrusion Alarm Systems Rev. 3 Not applicable N/A 10/1997 5.48 Design Considerations - Systems for Measuring 02/1975 Not applicable N/A the Mass of Liquids 5.49 Internal Transfers of Special Nuclear Material 03/1975 Not applicable N/A 5.51 Management Review of Nuclear Material 06/1975 Not applicable N/A Control and Accounting Systems 5.52 Standard Format and Content of a Licensee Rev. 3 Not applicable N/A Physical Protection Plan for Strategic Special 12/1994 Nuclear Material at Fixed Sites (Other than Nuclear Power Plants) 1.9-35 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (31 of 35)

Revision /

NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 5.53 Qualification, Calibration, and Error Estimation Rev. 1 Not applicable N/A Methods for Nondestructive Assay 02/1984 5.55 Standard Format and Content of Safeguards 03/1978 Not applicable N/A Contingency Plans for Fuel Cycle Facilities 5.56 Standard Format and Content of Safeguards 03/1978 Not applicable N/A Contingency Plans for Transportation 5.57 Shipping and Receiving Control of Strategic Rev. 1 Not applicable N/A Special Nuclear Material 06/1980 5.58 Considerations for Establishing Traceability of Rev. 1 Not applicable N/A Special Nuclear Material Accounting 02/1980 Measurements 5.59 Standard Format and Content for a Licensee Rev. 1 Not applicable N/A Physical Security Plan for the Protection of 02/1983 Special Nuclear Material of Moderate or Low Strategic Significance 5.60 Standard Format and Content of a Licensee 04/1980 Not applicable N/A Physical Protection Plan for Strategic Special Nuclear Material in Transit 5.61 Intent and Scope of the Physical Protection 06/1980 Not applicable N/A Upgrade Rule Requirements for Fixed Sites 5.62 Reporting of Safeguards Events Rev. 1 Not applicable N/A 11/1987 5.63 Physical Protection for Transient Shipments 07/1982 Not applicable N/A 5.65 Vital Area Access Controls, Protection of 09/1986 Not applicable (COL) N/A Physical Security Equipment, and Key and Lock Controls 1.9-36 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (32 of 35)

Revision /

NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 5.66 Access Authorization Program for Nuclear Rev. 2 Not applicable (COL) N/A Power Plants 10/2011 5.68 Protection Against Malevolent Use of Vehicles 08/1994 Not applicable (COL) N/A at Nuclear Power Plants 5.71 Cyber Security Programs for Nuclear Facilities 01/2010 Not applicable (COL) N/A 5.73 Fatigue Management for Nuclear Power Plant 03/2009 Not applicable (COL) N/A Personnel 5.74 Managing the Safety/Security Interface 06/2009 Not applicable (COL) N/A 5.75 Training and Qualification of Security Personnel 07/2009 Not applicable (COL) N/A at Nuclear Power Reactor Facilities 5.79 Protection of Safeguards Information 04/2011 Not applicable (COL) N/A 5.80 Pressure-Sensitive and Tamper-Indicating 12/2010 Not applicable (COL) N/A Device Seals for Material Control and Accounting of Special Nuclear Material 8.2 Administrative Practices in Radiation Surveys Rev. 1 Not applicable (COL) N/A and Monitoring 05/2011 8.4 Personal Monitoring Device-Direct Reading Rev. 1 Not applicable (COL) N/A Pocket Dosimeters 06/2011 8.7 Instructions for Recording and Reporting Rev. 2 Not applicable (COL) N/A Occupational Radiation Exposure Data 11/2005 1.9-37 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (33 of 35)

Revision /

NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 8.8 Information Relevant to Ensuring the Rev. 3 The APR1400 conforms with this NRC RG except for the Table 6.5-2, Occupational Radiation Exposures at Nuclear 06/1978 following. 10.4.6.1, 10.4.8.1.2, Power Stations will be ALARA

• Nickel-based alloy is used for SG tubes based on 11.2.1.2, 11.3.1.2, industry experience for similar applications in 11.3.1.3, 11.3.1.4, Korean domestic plants as described in Subsection 11.4.1.2, 11.4.1.3, 12.3.1.3. 11.4.2.4, 11.5.2.1, 11.5.2.2.3.1, 12.1.1.1, 12.1.2.1, 12.1.2.2, 12.1.3.2, 12.2.4, 12.3.1.1, 12.3.1.2, 12.3.1.3, 12.3.1.4, 12.3.1.5, 12.3.2.1, 12.3.2.3, 12.3.4 8.9 Acceptable Concepts, Models, Equations, and Rev. 1 Not applicable (COL) N/A Assumptions for a Bioassay Program 07/1993 8.10 Operating Philosophy for Maintaining Rev. 1-R Not applicable (COL) N/A Occupational Radiation Exposures As Low As Is 05/1977 Reasonably Achievable 8.11 Applications of Bioassay for Uranium 06/1974 Not applicable N/A 8.13 Instruction Concerning Prenatal Radiation Rev. 3 Not applicable (COL) N/A Exposure 06/1999 8.15 Acceptable Programs for Respiratory Protection Rev. 1 Not applicable (COL) N/A 10/1999 8.18 Information Relevant to Ensuring that Rev. 2 Not applicable (COL) N/A Occupational Radiation Exposures at Medical 04/2011 Institutions Will Be as Low as Reasonably Achievable 1.9-38 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (34 of 35)

Revision /

NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 8.19 Occupational Radiation Dose Assessment in Rev. 1 The APR1400 conforms with this NRC RG. 12.4.1.2 Light-Water Reactor Plants-Design Stage Man- 06/1979 Rem Estimates 8.20 Applications of Bioassay for I-125 and I-131 Rev. 1 Not applicable (COL) N/A 09/1979 8.21 Health Physics Surveys for Byproduct Material Rev. 1 Not applicable N/A at NRC-Licensed Processing and Manufacturing 10/1979 Plants 8.22 Bioassay at Uranium Mills Rev. 2 Not applicable N/A 05/2014 8.23 Radiation Safety Surveys at Medical Institutions Rev. 1 Not applicable N/A 01/1981 8.24 Health Physics Surveys During Enriched Rev. 2 Not applicable N/A Uranium-235 Processing and Fuel Fabrication 06/2012 8.25 Air Sampling in the Workplace Rev. 1 The APR1400 conforms with this NRC RG 12.3.1.4, 12.3.4 06/1992 8.26 Applications of Bioassay for Fission and 09/1980 Not applicable (COL) N/A Activation Products 8.27 Radiation Protection Training for Personnel at 03/1981 Not applicable (COL) N/A Light-Water-Cooled Nuclear Power Plants 8.28 Audible-Alarm Dosimeters 08/1981 Not applicable (COL) N/A 8.29 Instruction Concerning Risks from Occupational Rev. 1 Not applicable (COL) N/A Radiation Exposure 02/1996 8.30 Health Physics Surveys in Uranium Recovery Rev. 1 Not applicable N/A Facilities 05/2002 1.9-39 Rev. 2

APR1400 DCD TIER 2 Table 1.9-1 (35 of 35)

Revision /

NRC Regulatory Guide Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 8.31 Information Relevant to Ensuring that Rev. 1 Not applicable N/A Occupational Radiation Exposures at Uranium 05/2002 Recovery Facilities Will Be as Low as Is Reasonably Achievable 8.32 Criteria for Establishing a Tritium Bioassay 07/1988 Not applicable (COL) N/A Program 8.34 Monitoring Criteria and Methods To Calculate 07/1992 Not applicable (COL) N/A Occupational Radiation Doses 8.35 Planned Special Exposures Rev. 1 Not applicable (COL) N/A 08/2010 8.36 Radiation Dose to the Embryo/Fetus 07/1992 Not applicable (COL) N/A 8.37 ALARA Levels for Effluents from Materials 07/1993 Not applicable (COL) N/A Facilities 8.38 Control of Access to High and Very High Rev. 1 The APR1400 conforms with this NRC RG. Ch. 12 Radiation Areas of Nuclear Plants 05/2006 8.39 Release of Patients Administered Radioactive 04/1997 Not applicable N/A Materials 8.40 Methods for Measuring Effective Dose 07/2010 Not applicable (COL) N/A Equivalent from External Exposure 1.9-40 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (1 of 32)

APR1400 Conformance with the Standard Review Plan Revision /

SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 1.0 - Introduction and Interfaces Rev. 2 The APR1400 conforms with this SRP. Ch. 1 12/2011 2.0 - Site Characteristics and Site Parameters 03/2007 The APR1400 conforms with this SRP. Ch. 2 The site-specific data will be addressed in the combined license application (COLA).

2.1.1 - Site Location and Description Rev. 3 The APR1400 conforms with this SRP. 2.1.1 03/2007 The site-specific data will be addressed in the COLA.

2.1.2 - Exclusion Area Authority and Control Rev. 3 The APR1400 conforms with this SRP. 2.1.2 03/2007 The site-specific data will be addressed in the COLA.

2.1.3 - Population Distribution Rev. 3 The APR1400 conforms with this SRP. 2.1.3 03/2007 The site-specific data will be addressed in the COLA.

2.2.1 ~ 2.2.2 - Identification of Potential Hazards in Rev. 3 The APR1400 conforms with this SRP. 2.2.1~2.2.2 Site Vicinity 03/2007 The site-specific data will be addressed in the COLA.

2.2.3 - Evaluation of Potential Accidents Rev. 3 The APR1400 conforms with this SRP. 2.2.3 03/2007 The site-specific data will be addressed in the COLA.

2.3.1 - Regional Climatology Rev. 3 The APR1400 conforms with this SRP. 2.3.1 03/2007 The site-specific data will be addressed in the COLA.

2.3.2 - Local Meteorology Rev. 3 The APR1400 conforms with this SRP. 2.3.2 03/2007 The site-specific data will be addressed in the COLA.

2.3.3 - Onsite Meteorological Measurements Program Rev. 3 The APR1400 conforms with this SRP. 2.3.3 03/2007 The site-specific data will be addressed in the COLA.

2.3.4 - Short-Term Atmospheric Dispersion Estimates Rev. 3 The APR1400 conforms with this SRP. 2.3.4 for Accident Releases 03/2007 The site-specific data will be addressed in the COLA.

1.9-41 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (2 of 32)

Revision /

SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 2.3.5 - Long-Term Atmospheric Dispersion Estimates Rev. 3 The APR1400 conforms with this SRP. 2.3.5 for Routine Releases 03/2007 The site-specific data will be addressed in the COLA.

2.4.1 - Hydrologic Description Rev. 3 The APR1400 conforms with this SRP. 2.4.1 03/2007 The site-specific data will be addressed in the COLA.

2.4.2 - Floods Rev. 4 The APR1400 conforms with this SRP. 2.4.2 03/2007 The site-specific data will be addressed in the COLA.

2.4.3 - Probable Maximum Flood (PMF) on Streams Rev. 4 The APR1400 conforms with this SRP. 2.4.3 and Rivers 03/2007 The site-specific data will be addressed in the COLA.

2.4.4 - Potential Dam Failures Rev. 3 The APR1400 conforms with this SRP. 2.4.4 03/2007 The site-specific data will be addressed in the COLA.

2.4.5 - Probable Maximum Surge and Seiche Flooding Rev. 3 The APR1400 conforms with this SRP. 2.4.5 03/2007 The site-specific data will be addressed in the COLA.

2.4.6 - Probable Maximum Tsunami Hazards Rev. 3 The APR1400 conforms with this SRP. 2.4.6 03/2007 The site-specific data will be addressed in the COLA.

2.4.7 - Ice Effects Rev. 3 The APR1400 conforms with this SRP. 2.4.7 03/2007 The site-specific data will be addressed in the COLA.

2.4.8 - Cooling Water Canals and Reservoirs Rev. 3 The APR1400 conforms with this SRP. 2.4.8 03/2007 The site-specific data will be addressed in the COLA.

2.4.9 - Channel Diversions Rev. 3 The APR1400 conforms with this SRP. 2.4.9 03/2007 The site-specific data will be addressed in the COLA.

2.4.10 - Flooding Protection Requirements Rev. 3 The APR1400 conforms with this SRP. 2.4.10 03/2007 The site-specific data will be addressed in the COLA.

2.4.11 - Low Water Considerations Rev. 3 The APR1400 conforms with this SRP. 2.4.11 03/2007 The site-specific data will be addressed in the COLA.

1.9-42 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (3 of 32)

Revision /

SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 2.4.12 - Groundwater Rev. 3 The APR1400 conforms with this SRP. 2.4.12 03/2007 The site-specific data will be addressed in the COLA.

2.4.13 - Accidental Releases of Radioactive Liquid Rev. 3 The APR1400 conforms with this SRP. 2.4.13 Effluents in Ground and Surface Waters 03/2007 The site-specific data will be addressed in the COLA.

2.4.14 - Technical Specifications and Emergency Rev. 3 The APR1400 conforms with this SRP. 2.4.14 Operation Requirements 03/2007 The site-specific data will be addressed in the COLA.

2.5.1 - Basic Geologic and seismic Information Rev. 5 The APR1400 conforms with this SRP. 2.5.1 07/2014 The site-specific data will be addressed in the COLA.

2.5.2 - Vibratory Ground Motion Rev. 5 The APR1400 conforms with this SRP. 2.5.2 07/2014 The site-specific data will be addressed in the COLA.

2.5.3 - Surface Faulting Rev. 5 The APR1400 conforms with this SRP. 2.5.3 07/2014 The site-specific data will be addressed in the COLA.

2.5.4 - Stability of Subsurface Materials and Rev. 5 The APR1400 conforms with this SRP. 2.5.4 Foundations 07/2014 The site-specific data will be addressed in the COLA.

2.5.5 - Stability of Slopes Rev. 5 The APR1400 conforms with this SRP. 2.5.5 07/2014 The site-specific data will be addressed in the COLA.

3.2.1 - Seismic Classification Rev. 2 The APR1400 conforms with this SRP.

3.2.1 03/2007 3.2.2 - System Quality Group Classification Rev. 2 The APR1400 conforms with this SRP.

3.2.2 03/2007 3.3.1 - Wind Loading Rev. 3 The APR1400 conforms with this SRP.

3.3.1 03/2007 3.3.2 - Tornado Loads Rev. 3 The APR1400 conforms with this SRP. 3.3.2 03/2007 1.9-43 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (4 of 32)

Revision / DCD Tier 2 SRP Section/Title Issue Date Conformance or Summary Description of Deviation Section 3.4.1 - Internal Flood Protection for Onsite Rev. 3 The APR1400 conforms with this SRP. 3.4.1 Equipment Failures 03/2007 3.4.2 - Analysis Procedures Rev. 3 The APR1400 conforms with this SRP. 3.4.2 03/2007 3.5.1.1 - Internally Generated Missiles Rev. 3 The APR1400 conforms with this SRP. 3.5.1.1 (Outside Containment) 03/2007 3.5.1.2 - Internally-Generated Missiles Rev. 3 The APR1400 conforms with this SRP. 3.5.12 (Inside Containment) 03/2007 3.5.1.3 - Turbine Missiles Rev. 3 The APR1400 conforms with this SRP. 3.5.13 03/2007 3.5.1.4 - Missiles Generated by Tornadoes and Rev.4 (Draft) The APR1400 conforms with this SRP. 3.5.1.4 Extreme Winds 07/2013 3.5.1.5 - Site Proximity Missiles (Except Aircraft) Rev. 4 Not applicable (COL) N/A 03/2007 3.5.1.6 - Aircraft Hazards Rev. 4 Not applicable (COL) N/A 03/2010 3.5.2 - Structures, Systems, and Components to be Rev. 3 The APR1400 conforms with this SRP. 3.5.2 Protected from Externally-Generated Missiles 03/2007 3.5.3 - Barrier Design Procedures Rev. 3 The APR1400 conforms with this SRP. 3.5.3 03/2007 3.6.1 - Plant Design for Protection Against Postulated Rev. 3 The APR1400 conforms with this SRP. 3.6.1 Piping Failures in Fluid Systems Outside Containment 03/2007 3.6.2 - Determination of Rupture Locations and Rev. 3 The APR1400 conforms with this SRP. 3.6.2 Dynamic Effects Associated with the Postulated 12/2016 Rupture of Piping 1.9-44 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (5 of 32)

Revision /

SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 3.6.3 - Leak-Before-Break Evaluation Procedures Rev. 1 The APR1400 conforms with this SRP. 3.6.3 03/2007 3.7.1 - Seismic Design Parameters Rev.4 The APR1400 conforms with this SRP. 3.7.1 (Draft) 12/2012 3.7.2 - Seismic System Analysis Rev. 4 The APR1400 conforms with this SRP except for limits 3.7.2 09/2013 on response reduction due to incoherent seismic input motion.

3.7.3 - Seismic Subsystem Analysis Rev. 4 Alternate analysis methods are employed for piping 3.7.3 09/2013 systems.

No explicit range of the fundamental frequencies of components and equipment with respect to the dominant frequencies of the support structure is made.

3.7.4 - Seismic Instrumentation Rev. 2 The APR1400 conforms with this SRP. 3.7.4 03/2007 3.8.1 - Concrete Containment Rev. 4 The APR1400 conforms with this SRP. 3.8.1 09/2013 3.8.2 - Steel Containment Rev. 3 The APR1400 conforms with this SRP for the areas 3.8.2 05/2010 relating to Class MC steel portions of concrete containment.

3.8.3 - Concrete and Steel Internal Structures of Steel Rev. 4 The APR1400 conforms with this SRP. 3.8.3 or Concrete Containments 09/2013 3.8.4 - Other seismic Category I Structures Rev. 4 The APR1400 conforms with this SRP. 3.8.4 09/2013 3.8.5 - Foundations Rev. 4 The APR1400 conforms with this SRP. 3.8.5 09/2013 1.9-45 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (6 of 32)

Revision /

SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 3.9.1 - Special Topics for Mechanical Components Rev. 3 The APR1400 conforms with this SRP. 3.9.1 03/2007 3.9.2 - Dynamic Testing and Analysis of Systems, Rev. 3 The APR1400 conforms with this SRP with the following 3.9.2, App. 3.9B Structures, and Components 03/2007 exception:

• Startup testing with measurement of SG internals 3.9.3 - ASME Code Class 1, 2, and 3 Components and Rev. 3 The APR1400 conforms with this SRP. 3.9.3 Component Supports, and Core Support Structures 04/2014 3.9.4 - Control Rod Drive Systems Rev. 3 The APR1400 conforms with this SRP. 3.9.4 03/2007 3.9.5 - Reactor Pressure Vessel Internals Rev. 3 The APR1400 conforms with this SRP. 3.9.5 03/2007 3.9.6 - Functional Design, Qualification, and Inservice Rev. 3 The APR1400 conforms with this SRP. 3.9.6 Testing Programs for Pumps, Valves, and Dynamic 03/2007 Restraints 3.9.7 - Risk-Informed Inservice Testing 08/1998 The APR1400 conforms with this SRP. 3.9.7 3.9.8 - Risk-Informed Inservice Inspection of Piping 09/2003 The APR1400 conforms with this SRP. 3.9.8 3.10 - Seismic and Dynamic Qualification of Rev. 3 The APR1400 conforms with this SRP. 3.10 Mechanical and Electrical Equipment 03/2007 3.11 - Environmental Qualification of Mechanical and Rev. 3 The APR1400 conforms with this SRP. 3.11 Electrical Equipment 03/2007 3.12 - ASME Code Class 1, 2, and 3 Piping Systems, Rev. 1 The APR1400 conforms with this SRP. 3.12 Piping Components and Their Associated Supports 04/2014 3.13 - Threaded Fasteners-ASME Code Class 1, 2, 03/2007 The APR1400 conforms with this SRP. 3.13 and 3 1.9-46 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (7 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section BTP 3 Classification of Main Steam Components Rev. 2 Not applicable (BWRs only) N/A Other Than the Reactor Coolant Pressure Boundary for 03/2007 BWR Plants BTP 3 Classification of BWR/6 Main Steam and Rev. 2 Not applicable (BWRs only) N/A Feedwater Components Other Than the Reactor 03/2007 Coolant Pressure Boundary BTP 3 Protection against Postulated Piping Failures Rev. 3 The APR1400 conforms with this BTP. 3.6.1, 10.4.4.3 in Fluid Systems Outside Containment 03/2007 BTP 3 Postulated Rupture Locations in Fluid Rev. 2 The APR1400 conforms with this BTP. 3.6.2, 10.4.4.3 System Piping Inside and Outside Containment 03/2007 4.2 - Fuel System Design Rev. 3 The APR1400 conforms with this SRP. 4.2 03/2007 4.3 - Nuclear Design Rev. 3 The APR1400 conforms with this SRP. 4.3 03/2007 4.4 - Thermal and Hydraulic Design Rev. 2 The APR1400 conforms with this SRP. 4.4 03/2007 4.5.1 - Control Rod Drive Structural Materials Rev. 3 The APR1400 conforms with this SRP except for the 4.5.1 03/2007 following:

• The usage of control drive structural material with a yield strength greater than 90 ksi is limited to the steel ball in the vent valve on the top of the CEDMs, bearing inserts, and alignment tab in the motor assembly.

4.5.2 - Reactor Internal and Core Support Structure Rev. 3 The APR1400 conforms with this SRP. 4.5.2 Materials 03/2007 1.9-47 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (8 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 4.6 - Functional Design of Control Rod Drive System Rev. 2 The APR1400 conforms with this SRP. 3.6, 3.9.4, 4.6, 6.3, 03/2007 7.2, 7.7, 9.3.4, 9.4, Chapter 14, Chapter 15 BTP 4 Westinghouse Constant Axial Offset Control Rev. 3 Not applicable N/A (CAOC) 03/2007 5.2.1.1 - Compliance with the Codes and Standards Rev. 3 The APR1400 conforms with this SRP. 5.2.1.1 Rule, 10 CFR 50.55a 03/2007 5.2.1.2 - Applicable Code Cases Rev. 3 The APR1400 conforms with this SRP. 5.2.1.2 03/2007 5.2.2 - Overpressure Protection Rev. 3 The APR1400 conforms with this SRP. 5.2.2 03/2007 5.2.3 - Reactor Coolant Pressure Boundary Materials Rev. 3 The APR1400 conforms with this SRP except for the 5.2.3 03/2007 following:

The electroslag weld process is not used in the fabrication of any RCPB components.

5.2.4 - Reactor Coolant Pressure Boundary Inservice Rev. 2 The APR1400 conforms with this SRP. 5.2.4 Inspection and Testing 03/2007 5.2.5 - Reactor Coolant Pressure Boundary Leakage Rev. 2 The APR1400 conforms with this SRP. 5.2.5 Detection 03/2007 5.3.1 - Reactor Vessel Materials Rev. 2 The APR1400 conforms with this SRP except for the 5.3.1 03/2007 following:

Actual reactor vessel materials are tested at the time of material procurement. Test requirements are described in Subsection 5.3.1.5.

5.3.2 - Pressure-Temperature Limits, Upper-Shelf Rev. 2 The APR1400 conforms with this SRP. 5.3.2 Energy, and Pressurized Thermal Shock 03/2007 1.9-48 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (9 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 5.3.3 - Reactor Vessel Integrity Rev. 2 The APR1400 conforms with this SRP. 5.3.3 03/2007 5.4 - Reactor Coolant System Component and Rev. 2 The APR1400 conforms with this SRP. 5.4 Subsystem Design 03/2007 5.4.1.1 - Pump Flywheel Integrity (PWR) Rev. 3 The APR1400 conforms with this SRP. 5.4.1.1 05/2010 5.4.2.1 - Steam Generator Materials Rev. 3 The APR1400 conforms with this SRP. 5.4.2.1 03/2007 5.4.2.2 - Steam Generator Program Rev. 2 The APR1400 conforms with this SRP. 5.4.2.2 03/2007 5.4.6 - Reactor Core Isolation Cooling System (BWR) Rev. 4 Not applicable (BWRs only) N/A 03/2007 5.4.7 - Residual Heat Removal (RHR) System Rev. 5 The APR1400 conforms with this SRP. 5.4.7 05/2010 5.4.8 - Reactor Water Cleanup System (BWR) Rev. 3 Not applicable (BWRs only) N/A 03/2007 5.4.11 - Pressurizer Relief Tank Rev. 4 The APR1400 conforms with this SRP. 5.4.11 05/2010 5.4.12 - Reactor Coolant System High Point Vents Rev. 1 The APR1400 conforms with this SRP. 5.4.12 03/2007 5.4.13 - Isolation Condenser System (BWR) 03/2007 Not applicable (BWRs only) N/A BTP 5 Monitoring of Secondary Side Water Rev. 3 The APR1400 conforms with this BTP. 10.4.8.3 Chemistry in PWR Steam Generators 03/2007 1.9-49 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (10 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section BTP 5 Overpressurization Protection of Rev. 3 The APR1400 conforms with this BTP. 5.2.2.1.1 Pressurized-Water Reactors While Operating at Low 03/2007 5.2.2.1.2 Temperatures BTP 5 Fracture Toughness Requirements Rev. 2 The APR1400 conforms with this BTP. 5.2.3.3, 5.3.1.1, 03/2007 5.3.1.5, 5.3.1.6, 5.3.2, 5.3.2.3, 5.3.2.4 BTP 5 Design Requirements of the Residual Heat Rev. 4 The APR1400 conforms with this BTP. 5.4.7.1.2, 5.2.2.1.1 Removal System 03/2007 6.1.1 - Engineered Safety Features Materials Rev. 2 The APR1400 conforms with this SRP. 6.1.1 03/2007 6.1.2 - Protective Coating Systems (Paints)-Organic Rev. 3 The APR1400 conforms with this SRP. 6.1.2 Materials 03/2007 6.2.1 - Containment Functional Design Rev. 3 The APR1400 conforms with this SRP. 6.2.1 03/2007 6.2.1.1.A - PWR Dry Containments, Including Rev. 3 The APR1400 Conforms with this SRP. 6.2.1 Subatmospheric Containments 03/2007 As for Criterion 9, the structural design pressure of each subcompartment is determined based on the design experience.

6.2.1.1.B - Ice Condenser Containments Rev. 2 Not applicable N/A 07/1981 6.2.1.1.C - Pressure-Suppression Type BWR Rev. 7 Not applicable (BWR) N/A Containments 03/2007 6.2.1.2 - Subcompartment Analysis Rev. 3 The APR1400 conforms with this SRP. 6.2.1.2 03/2007 1.9-50 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (11 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 6.2.1.3 - Mass and Energy Release Analysis for Rev. 3 The APR1400 conforms with this SRP except for the 6.2.1.3 Postulated Loss-of-Coolant Accidents (LOCAs) 03/2007 following:

Metal-water reaction energy is not included in the mass/energy source terms since this energy has been shown to have a small effect on the containment pressure.

6.2.1.4 - Mass and Energy Release Analysis for Rev. 2 The APR1400 conforms with this SRP. 6.2.1.4 Postulated Secondary System Pipe Ruptures 03/2007 6.2.1.5 - Minimum Containment Pressure Analysis for Rev. 3 The APR1400 conforms with this SRP. 6.2.1.5 Emergency Core Cooling System Performance 03/2007 Capability Studies 6.2.2 - Containment Heat Removal Systems Rev. 5 Conformance with exceptions. Criterion 4 is not applied 6.2.2 03/2007 to APR1400, because the APR1400 does not have the fan cooler system for containment heat removal following the design base accident.

6.2.3 - Secondary Containment Functional Design Rev. 3 Not applicable N/A 03/2007 The APR1400 does not have a secondary containment.

6.2.4 - Containment Isolation System Rev. 3 The APR1400 conforms with this SRP. 6.2.4 03/2007 The Chapter 15 dose analysis showed the acceptability of 30-second closure times for the purge valves.

6.2.5 -Combustible Gas Control in Containment Rev. 3 The APR1400 conforms with this SRP. 6.2.5 03/2007 6.2.6 - Containment Leakage Testing Rev. 3 The APR1400 conforms with this SRP. 6.2.6 03/2007 6.2.7 - Fracture Prevention of Containment Pressure Rev. 1 The APR1400 conforms with this SRP. 6.2.7 Boundary 03/2007 1.9-51 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (12 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 6.3 - Emergency Core Cooling System Rev. 3 The APR1400 conforms with exceptions relevant to 6.3 03/2007 criteria applied to BWRs.

BTP 6-5 Item E is applied to traditional PWR with a switchover from the injection mode to the recirculation cooling mode.

6.4 - Control Room Habitability System Rev. 3 The APR1400 conforms with this SRP except for the following. 6.4 03/2007 The control room habitability during a postulated hazardous chemical release is addressed in COLA.

6.5.1 - Engineered Safety Features (ESF) Atmosphere Rev. 4 The APR1400 conforms with this SRP. 6.5.1 Cleanup Systems 05/2010 6.5.2 - Containment Spray as a Fission Product Rev. 4 The APR1400 conforms with this SRP. 6.5.2 Cleanup System 03/2007 Conformance with exceptions. Criterion 3B is not applied to APR1400, because the APR1400 does not have the containment spray chemical additive tanks.

6.5.3 - Fission Product Control Systems and Structures Rev. 3 The APR1400 analysis assumes more than 50% mixing. 6.5.3 03/2007 6.5.4 - Ice Condenser as a Fission Product Cleanup Rev. 3 Not applicable N/A System 12/1988 6.5.5 - Pressure Suppression Pool as a Fission Product Rev. 1 Not applicable (BWRs only) N/A Cleanup System 03/2007 The APR1400 has adopted the Containment Spray System as Fission Product Cleanup System to control the fission products released during postulated reactor accidents 6.6 - Inservice Inspection and Testing of Class 2 and 3 Rev. 2 The APR1400 conforms with this SRP. 6.6 Components 03/2007 6.7 - Main Steam Isolation Valve Leakage Control Rev. 2 Not applicable (BWR) N/A System (BWR) 07/1981 BTP 6 pH for Emergency Coolant Water for 03/2007 The APR1400 conforms with this BTP. 6.5.2 Pressurized Water Reactors 1.9-52 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (13 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section BTP 6 Minimum Containment Pressure model for Rev. 3 The APR1400 conforms with this BTP. 6.2 PWR ECCS Performance Evaluation 03/2007 BTP 6 Determination of Bypass Leakage Paths in Rev. 3 Not applicable N/A Dual Containment Plants 03/2007 The APR1400 does not have a dual containment.

BTP 6 Containment Purging during Normal Plant Rev. 3 The APR1400 conforms with this BTP. 9.4.6.2 Operations 03/2007 BTP 6 Currently the Responsibility of Reactor Rev. 3 The APR1400 design conforms with this BTP except for 6.3.2.5.2 Systems Piping from the RWST (or BWST) and 03/2007 the following:

Containment Sump(s) to the Safety Injection Pumps The APR1400 adopted the IRWST (In-containment Refueling Water Storage Tank) design feature.

Therefore, containment sumps and the recirculation mode concept are not applied in the APR1400 design.

7.0 - Instrumentation and Controls - Overview of Rev. 6 The APR1400 conforms with this SRP. 7.1 Review Process 05/2010 App. 7.0-A - Review Process for Digital Rev. 5 The APR1400 conforms with this SRP. 7.1 Instrumentation and Control Systems 03/2007 7.1 - Instrumentation and Controls - Introduction Rev. 5 The APR1400 conforms with this SRP. 7.1 03/2007 7.1-T - Table 7-1 Regulatory Requirements, Rev. 5 The APR1400 conforms with this SRP. 7.1 Acceptance Criteria, and Guidelines for 03/2007 Instrumentation and Control Systems Important to Safety App. 7.1-A - Acceptance Criteria and Guidelines for Rev. 5 The APR1400 conforms with this SRP. 7.1 Instrumentation and Control Systems Important to 03/2007 Safety 1.9-53 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (14 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section App. 7.1-B - Guidance for Evaluation of Conformance Rev. 5 Not applicable N/A to IEEE Std. 279 03/2007 App. 7.1-C - Guidance for Evaluation of Conformance Rev. 5 The APR1400 conforms with this SRP. 7.1 to IEEE Std. 603 03/2007 App. 7.1-D - Guidance for Evaluation of the 03/2007 The APR1400 conforms with this SRP. 7.1 Application of IEEE Std. 7-4.3.2 7.2 - Reactor Trip System Rev. 5 The APR1400 conforms with this SRP. 7.2 03/2007 7.3 - Engineered Safety Features Systems Rev. 5 The APR1400 conforms with this SRP. 7.3 03/2007 7.4 - Safe Shutdown Systems Rev. 5 The APR1400 conforms with this SRP. 7.4 03/2007 7.5 - Information Systems Important to Safety Rev. 5 The APR1400 conforms with this SRP. 7.5 03/2007 7.6 - Interlock Systems Important to Safety Rev. 5 The APR1400 conforms with this SRP except for the 7.6 03/2007 following:

Interlocks for shutdown cooling system (SCS) suction isolation valves are not diverse.

7.7 - Control Systems Rev. 5 The APR1400 conforms with this SRP. 7.7 03/2007 7.8 - Diverse Instrumentation and Control Systems Rev. 5 The APR1400 conforms with this SRP. 7.8 03/2007 7.9 - Data Communication Systems Rev. 5 The APR1400 conforms with this SRP. 7.9 03/2007 1.9-54 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (15 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section BTP 7 Guidance on Isolation of Low-Pressure Rev. 5 The APR1400 conforms with this BTP. 7.1.2.58, Systems from the High-Pressure Reactor Coolant 03/2007 Table 7.1-1 System BTP 7 Guidance on Requirements of Motor- Rev. 5 The APR1400 conforms with this BTP. 7.1.2.59, Operated Valves in the Emergency Core Cooling 03/2007 Table 7.1-1 System Accumulator Lines BTP 7 Guidance on Protection System Trip Point Rev. 5 Not applicable N/A Changes for Operation With Reactor Coolant Pumps 03/2007 Out of Service BTP 7 Guidance on Design Criteria for Auxiliary Rev. 5 The APR1400 conforms with this BTP. 7.1.2.61, Feedwater Systems 03/2007 Table 7.1-1 BTP 7 Guidance on Spurious Withdrawals of Rev. 5 The APR1400 conforms with this BTP. 7.1.2.62, Single Control Rods in Pressurized Water Reactors 03/2007 Table 7.1-1 BTP 7 Guidance on Design of Instrumentation and Rev. 5 The APR1400 adopted the in-containment refueling water N/A Controls Provided to Accomplish Changeover from 03/2007 storage tank (IRWST). Therefore, containment sumps and Injection to Recirculation Mode the recirculation mode concept are not applied in the APR1400.

BTP 7 Guidance for Application of Regulatory Rev. 5 The APR1400 conforms with this BTP. 7.1.2.64, Guide 1.22 03/2007 Table 7.1-1 BTP 7 Guidance on Requirements for Reactor Rev. 5 The APR1400 conforms with this BTP. 7.1.2.65, Protection System Anticipatory Trips 03/2007 Table 7.1-1 BTP 7 Guidance on Application of Regulatory Rev. 5 The APR1400 conforms with this BTP. 7.1.2.66, Guide 1.97 03/2007 Table 7.1-1 1.9-55 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (16 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section BTP 7 Guidance on Application and Qualification Rev. 5 The APR1400 conforms with this BTP. 7.1.2.67, of Isolation Devices 03/2007 Table 7.1-1 BTP 7 Guidance on Establishing and Maintaining Rev. 5 The APR1400 conforms with this BTP except for 7.1.2.68, Instrument Setpoints 03/2007 acceptance Criterion 4. Table 7.1-1 For the APR1400, surveillance and calibration interval is set by 18-months as considering overhaul interval.

BTP 7 Guidance on Cross-Calibration of Rev. 5 The APR1400 conforms with this BTP. 7.1.2.69, Protection System Resistance Temperature Detectors 03/2007 Table 7.1-1 BTP 7 Guidance on Software Reviews for Digital Rev. 5 The APR1400 conforms with this BTP. 7.1.2.70, Computer-Based Instrumentation and Controls Systems 03/2007 Table 7.1-1 BTP 7 Guidance on Self-Test and Surveillance Rev. 5 The APR1400 conforms with this BTP. 7.1.2.71, Test Provisions 03/2007 Table 7.1-1 BTP 7 Guidance on the Use of Programmable Rev. 5 The APR1400 conforms with this BTP. 7.1.2.72, Logic Controllers in Digital Computer-Based 03/2007 Table 7.1-1 Instrumentation and Control Systems Rev. 6 The APR1400 conforms with this BTP. 7.1.2.73, BTP 7 Guidance for Evaluation of Diversity and 07/2012 Table 7.1-1, 7.3.2.4, Defense-In-Depth in Digital Computer-Based 7.8.2.1 Instrumentation and Control Systems 7.8.2.2, 7.8.2.3, 7.8.3.1 BTP 7 Guidance on Digital Computer Real-Time Rev. 5 The APR1400 conforms with this BTP. 7.1.2.74, Performance 03/2007 Table 7.1-1 1.9-56 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (17 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 8.1 - Electric Power - Introduction Rev. 4 The APR1400 conforms with this SRP. 8.1 02/2012 8.2 - Offsite Power System Rev. 5 The APR1400 conforms with this SRP. 8.2 05/2010 8.3.1 - AC Power Systems (Onsite) Rev. 4 The APR1400 conforms with this SRP. 8.3.1 05/2010 8.3.2 - DC Power Systems (Onsite) Rev. 4 The APR1400 conforms with this SRP. 8.3.2 05/2010 8.4 - Station Blackout Rev. 1 The APR1400 conforms with this SRP. 8.4 05/2010 App. 8-A - General Agenda, Station Site Visits Rev. 1 Not applicable (COL) N/A 03/2007 BTP 8 Requirements on Motor-Operated Valves in Rev. 3 The APR1400 conforms with this BTP. 8.1.3.3, 8.3.1.2.4, the ECCS Accumulator Lines 03/2007 Table 8.1-2 BTP 8 Use of Diesel Generator Sets for Peaking Rev. 3 The emergency diesel generator (EDG) provides backup 8.1.3.3, 8.3.1.2.4, 03/2007 power to the safety-related loads for safety shutdown Table 8.1-2 during a loss of offsite power (LOOP). However, the EDG is not used for peaking service for offsite power system. The APR1400 conforms with this BTP.

BTP 8 Stability of Offsite Power Systems Rev. 3 Not applicable (COL) N/A 03/2007 BTP 8 Application of the Single Failure Criterion Rev. 3 The APR1400 conforms with this BTP. 8.1.3.3, 8.3.1.2.4 to Manually Controlled Electrically Operated Valves 03/2007 Table 8.1-2 1.9-57 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (18 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section BTP 8 Supplemental Guidance for Bypass and Rev. 3 The APR1400 conforms with this BTP. 8.1.3.3, 8.3.1.2.4, Inoperable Status Indication for Engineered Safety 03/2007 Table 8.1-2 Features Systems BTP 8 Adequacy of Station Electric Distribution Rev. 3 The APR1400 conforms with this BTP 8.1.3.3, 8.2.2.3, System Voltages 03/2007 8.3.1.1.2.3, 8.3.1.1.3.11, 8.3.1.3.4, Table 8.1-2 BTP 8 Criteria for Alarms and Indications Rev. 3 The APR1400 conforms with this BTP. 7.5.1.3, 8.1.3.3, Associated with Diesel-Generator Unit Bypassed and 03/2007 8.3.1.2.4, Inoperable Status Table 8.1-2 BTP 8 Onsite (Emergency Diesel Generators) and 02/2012 Not applicable N/A Offsite Power Sources Allowed Outage Time Extensions BTP 8 Open Phase Conditions in Electric Rev. 0 The APR1400 conforms with this BTP 8.1.3.1, 8.1.3.3, Power System 07/2015 8.2.1.3, 8.2.2.3, Table 8.1-2 9.1.1 - Criticality Safety of Fresh and Spent Fuel Rev. 3 The APR1400 conforms with this SRP. 9.1.1 Storage and Handling 03/2007 9.1.2 - New and Spent Fuel Storage Rev. 4 The APR1400 conforms with this SRP. 9.1.2 03/2007 9.1.3 - Spent Fuel Pool Cooling and Cleanup System Rev. 2 The APR1400 conforms with this SRP. 9.1.3 03/2007 1.9-58 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (19 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 9.1.4 - Light Load Handling System and Refueling Rev. 4 The APR1400 conformance with acceptance criteria 5 is 9.1.4 Cavity 07/2014 not applicable for the APR1400 design certification.

(APR1400 is a single unit.)

9.1.5 - Overhead Heavy Load Handling Systems Rev. 1 The APR1400 conformance with exceptions. 9.1.5 03/2007 Criterion 5 is not applicable for the APR1400 design certification. (APR1400 is a single unit.)

9.2.1 - Station Service Water System Rev. 5 The APR1400 conforms with this SRP. 9.2.1 03/2007 9.2.2 - Reactor Auxiliary Cooling Water System Rev. 4 The APR1400 conforms with this SRP. 9.2.2 03/2007 9.2.4 - Potable and Sanitary Water Systems Rev. 3 The APR1400 conforms with this SRP. 9.2.4 03/2007 9.2.5 - Ultimate Heat Sink Rev. 3 The APR1400 conforms with this SRP. 9.2.5 03/2007 9.2.6 - Condensate Storage Facilities Rev. 3 The APR1400 conforms with this SRP with exceptions. 9.2.6 03/2007 Since the condensate storage facilities have no safety-related function and the APR1400 is not multi-unit, Criteria 2,3,4,5,6, and 7 of this SRP are not applicable.

9.3.1 - Compressed Air System Rev. 2 The APR1400 conforms with this SRP. 9.3.1 03/2007 9.3.2 - Process and Post-Accident Sampling Systems Rev. 3 The APR1400 conforms with this SRP. 9.3.2 03/2007 9.3.3 - Equipment and Floor Drainage System Rev. 3 The APR1400 conforms with this SRP. 9.3.3 03/2007 1.9-59 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (20 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 9.3.4 - Chemical and Volume Control System (PWR) Rev. 3 The APR1400 conforms with this SRP. 9.3.4 (Including Boron Recovery System) 03/2007 9.3.5 - Standby Liquid Control System (BWR) Rev. 3 Not applicable (BWR) N/A 03/2007 9.4.1 - Control Room Area Ventilation System Rev. 3 The APR1400 conforms with this SRP. 9.4.1 03/2007 9.4.2 - Spent Fuel Pool Area Ventilation System Rev. 3 The APR1400 conformance with exceptions. Criterion 2 9.4.2 03/2007 is not applicable (Not multiple unit plants) 9.4.3 - Auxiliary and Radwaste Area Ventilation Rev. 3 The APR1400 conforms with this SRP. 9.4.3, 9.4.7 System 03/2007 9.4.4 - Turbine Area Ventilation System Rev. 3 The APR1400 conforms with this SRP. 9.4.4 03/2007 9.4.5 - Engineered Safety Feature Ventilation System Rev. 3 The APR1400 conformance with exceptions. 9.4.5 03/2007 Criterion 3: Not multiple unit plants.

Criterion 5: Air cleanup function is provided for aux.

building controlled area HVAC system only.

9.5.1.1 - Fire Protection Program 02/2009 The APR1400 conforms with this SRP. 9.5.1 9.5.1.2 - Risk-Informed (RI), Performance-Based (PB) 12/2009 Not applicable. N/A Fire Protection Program (FPP) This SRP is allowed to the operating nuclear power reactor licensees to adopt risk-informed, performance-based approach as an alternative to the existing deterministic fire protection requirement.

APR 1400 fire protection is designed to the requirements of SRP 9.5.1.1 which provides deterministic fire protection guidance, it is not necessary to incorporate this SRP.

1.9-60 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (21 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 9.5.2 - Communications Systems Rev. 3 The APR1400 conforms with exceptions. 9.5.2 03/2007 Acceptance criteria 1, 2, 3, 12, 13, and 14 refer to site-specific emergency response and security requirements that are the responsibility of the COL applicant.

9.5.3 - Lighting Systems Rev. 3 The APR1400 conforms with this SRP. 9.5.3 03/2007 9.5.4 - Emergency Diesel Engine Fuel Oil Storage and Rev. 3 The APR1400 conforms with this SRP. 9.5.4 Transfer System 03/2007 9.5.5 - Emergency Diesel Engine Cooling Water Rev. 3 The APR1400 conforms with this SRP. 9.5.5 System 03/2007 9.5.6 - Emergency Diesel Engine Starting System Rev. 3 The APR1400 conforms with this SRP. 9.5.6 03/2007 9.5.7 - Emergency Diesel Engine Lubrication System Rev. 3 The APR1400 conforms with this SRP. 9.5.7 03/2007 9.5.8 - Emergency Diesel Engine Combustion Air Rev. 3 The APR1400 conforms with this SRP. 9.5.8 Intake and Exhaust System 03/2007 10.2 - Turbine Generator Rev. 3 The APR1400 conformance with exceptions. 10.2 03/2007 SRP 10.2 II. Acceptance Criteria 3 is not applicable because there is no safety-related equipment in the turbine room.

1.9-61 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (22 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 10.2.3 - Turbine Rotor Integrity Rev. 2 The APR1400 conforms with this SRP except FATT and 10.2.3 03/2007 Charpy V-notch energy in the material selection.

The values specified in SRP 10.2.3.II.1 are based on material acceptance data taken from specimens at the surface of a shrunk-on wheel forgings. The values in the APR1400 DCD Tier 2 are specified that FATT is no higher than -1 °C (30 °F) and Cv energy is at least 6.22 kg-m (45 ft-lbs). These values are based on deep-seated specimens near the center of the integral rotor forging.

(The material testing has shown that FATT increases from the outer surface to the deep-seated region of the forging as a result of variation in the cooling rate during the quenching process.)

10.3 - Main Steam Supply System Rev. 4 The APR1400 conforms with this SRP. 10.3 03/2007 10.3.6 - Steam and Feedwater System Materials Rev. 3 The APR1400 conforms with this SRP. 10.3.6 03/2007 10.4.1 - Main Condensers Rev. 3 The APR1400 conforms with this SRP. 10.4.1 03/2007 10.4.2 - Main Condenser Evacuation System Rev. 3 The APR1400 conformance with exceptions. 10.4.2 03/2007 Criterion 1 refers to a potential for explosive mixtures and the APR1400 has no potential for explosive mixtures.

10.4.3 - Turbine Steam Seal System Rev. 3 The APR1400 conforms with this SRP. 10.4.3 03/2007 1.9-62 Rev. 2

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 10.4.4 - Turbine Bypass System Rev. 3 The APR1400 conforms with this SRP. 10.4.4 03/2007 10.4.5 - Circulating Water System Rev. 3 System is site-specific and is addressed with interface 10.4.5 03/2007 requirements.

10.4.6 - Condensate Cleanup System Rev. 3 The APR1400 conforms with this SRP. 10.4.6 03/2007 10.4.7 - Condensate and Feedwater System Rev. 4 The APR1400 conformance with exceptions. 10.4.7 03/2007 Criterion 3 refers to shared systems and the APR1400 is a single unit design.

Criterion 7 is defined as COL item in DCD subsection 10.3.6.

Criterion 8 is for BWR.

10.4.8 - Steam Generator Blowdown System (PWR) Rev. 3 The APR1400 conforms with this SRP. 10.4.8 03/2007 10.4.9 - Auxiliary Feedwater System (PWR) Rev. 3 The APR1400 conforms with this SRP. 10.4.9 03/2007 BTP 10 Design Guidelines for Auxiliary Feedwater Rev. 3 The APR1400 conforms with this BTP. 10.4.9 System Pump Drive and Power Supply Diversity for 03/2007 Pressurized Water Reactor Plants Rev. 4 The APR1400 conforms with this BTP. 10.4.7.6, BTP 10 Design Guidelines for Avoiding Water 03/2007 10.4.9.1.2, Hammers in Steam Generators 10.4.9.3 11.1 - Source Terms Rev. 3 The APR1400 conforms with this SRP. 11.1 03/2007 1.9-63 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (24 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 11.2 - Liquid Waste Management Systems Rev. 4 The APR1400 conforms with this SRP. 11.2 05/2010 However, cost-benefit analysis for liquid waste management systems is deferred to the site-specific application due to the site-specific nature of population dose analyses.

The plant transients that might occur less frequently than once per fuel cycle are not taken into account for the design of waste collection tanks and waste sample tanks.

11.3 - Gaseous Waste Management System Rev. 3 The APR1400 conforms with this SRP. 11.3 03/2007 However, cost-benefit analysis for gaseous waste management systems is deferred to the site-specific application.

11.4 - Solid Waste Management System Rev. 3 The APR1400 conforms with this SRP. 11.4 03/2007 However, cost-benefit analysis for gaseous waste management systems is deferred to the site-specific application.

11.5 - Process and Effluent Radiological Monitoring Rev. 5 The APR1400 conforms with this SRP. 11.5 Instrumentation and Sampling Systems 05/2010 BTP 11 Design Guidance for Solid Radioactive Rev. 3 The APR1400 conforms with this BTP. 11.4 Waste Management Systems Installed in Light-Water- 03/2007 Cooled Nuclear Power Reactor Plants BTP 11 Postulated Radioactive Releases Due to a Rev. 3 The APR1400 conforms with this BTP. 11.3.3.2 Waste Gas System Leak or Failure 03/2007 BTP 11 Postulated Radioactive Releases Due to Rev. 3 The APR1400 conforms with this BTP except for the site- 11.2.3.2 Liquid-Containing Tank Failures 03/2007 specific features. The COL applicant is to provide the site-specific geology and hydrology.

1.9-64 Rev. 2

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 12.1 - Assuring that Occupational Radiation Exposures Rev. 4 The APR1400 conforms with this SRP except for the 12.1 Are As Low As Is Reasonably Achievable 09/2013 acceptance Criteria 3 and 4. These criteria are addressed in the COLA.

12.2 - Radiation Sources Rev. 4 The APR1400 conforms with this SRP. 12.2 09/2013 12.3-12.4 - Radiation Protection Design Features Rev. 5 The APR1400 conforms with this SRP. 12.3, 12.4 09/2013 12.5 - Operational Radiation Protection Program Rev. 5 Not applicable (COL) N/A 09/2013 13.1.1 - Management and Technical Support Rev. 5 Not applicable (COL) N/A Organization 03/2007 13.1.2-13.1.3 - Operating Organization Rev. 6 Not applicable (COL) N/A 03/2007 13.2.1 - Reactor Operator Requalification Program; Rev. 3 Not applicable (COL) N/A Reactor Operator Training 03/2007 13.2.2 - Non-Licensed Plant Staff Training Rev. 3 Not applicable (COL) N/A 03/2007 13.3 - Emergency Planning Rev. 3 Conformance with an exception. 13.3 03/2007 Under COL applicants responsibility.

The design feature, facilities, functions, and equipment necessary for emergency planning are included in DCD Tier 2.

13.4 - Operational Programs Rev. 3 Not applicable (COL) N/A 03/2007 13.5.1.1 - Administrative Procedures-General Rev.1 Not applicable (COL) N/A 12/2011 1.9-65 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (26 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 13.5.2.1 - Operating and Emergency Operating Rev. 2 Not applicable (COL) N/A Procedures 03/2007 13.6 - Physical Security Rev. 3 The APR1400 conforms with this SRP. 13.6 03/2007 13.6.1 - Physical Security - Combined License and Rev. 1 Not applicable (COL) N/A Operating Reactors 10/2010 13.6.2 - Physical Security - Design Certification Rev. 1 The APR1400 conforms with this SRP. 13.6.2 10/2010 13.6.3 - Physical Security - Early Site Permit Rev. 1 Not applicable (COL) N/A 10/2010 13.6.6 - Cyber Security Plan 11/2010 Not applicable (COL) N/A 14.2 - Initial Plant Test Program - Design Certification Rev. 3 The APR1400 conforms with this SRP. 14.2 and New License Applicants 03/2007 14.2.1 - Generic Guidelines for Extended Power 08/2006 Not applicable N/A Uprate Testing Programs 14.3 - Inspections, Tests, Analyses, and Acceptance 03/2007 The APR1400 conforms with this SRP. 14.3 Criteria 14.3.2 - Structural and Systems Engineering - 03/2007 The APR1400 conforms with this SRP. 14.3.2.2 Inspections, Tests, Analyses, and Acceptance Criteria 14.3.3 - Piping Systems and Components - 03/2007 The APR1400 conforms with this SRP. 14.3.2.3 Inspections, Tests, Analyses, and Acceptance Criteria 14.3.4 - Reactor Systems - Inspections, Tests, 03/2007 The APR1400 conforms with this SRP. 14.3.2.4 Analyses, and Acceptance Criteria 14.3.5 - Instrumentation and Controls - Inspections, 03/2007 The APR1400 conforms with this SRP. 14.3.2.5 Tests, Analyses, and Acceptance Criteria 1.9-66 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (27 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 14.3.6 - Electrical Systems - Inspections, Tests, 03/2007 The APR1400 conforms with this SRP. 14.3.2.6 Analyses, and Acceptance Criteria 14.3.7 - Plant Systems - Inspections, Tests, Analyses, 03/2007 The APR1400 conforms with this SRP. 14.3.2.7 and Acceptance Criteria 14.3.8 - Radiation Protection - Inspections, Tests, 03/2007 The APR1400 conforms with this SRP. 14.3.2.8 Analyses, and Acceptance Criteria 14.3.9 - Human Factors Engineering - Inspections, 03/2007 The APR1400 conforms with this SRP. 14.3.2.9 Tests, Analyses, and Acceptance Criteria 18.4 ~ 18.12 14.3.10 - Emergency Planning - Inspections, Tests, 03/2007 The APR1400 conforms with this SRP. 14.3.2.10 Analyses, and Acceptance Criteria 18.4 ~ 18.12 14.3.11 - Containment Systems - Inspections, Tests, 03/2007 The APR1400 conforms with this SRP. 14.3.2.11 Analyses, and Acceptance Criteria 14.3.12 - Physical Security Hardware - Inspections, Rev. 1 The APR1400 conforms with this SRP. 14.3.2.12 Tests, Analyses, and Acceptance Criteria 05/2010 15.0 - Introduction - Transient and Accident Analyses Rev. 3 The APR1400 conforms with this SRP. 15.0 03/2007 15.0.1 - Radiological Consequence Analyses Using 07/2000 Not applicable. N/A Alternative Source Terms This SRP applies to operating plants adopting alternative source term inputs.

15.0.2 - Review of Transient and Accident Analysis 03/2007 The APR1400 conforms with this SRP. 15.02 Method 15.0.3 - Design Basis Accident Radiological 03/2007 The APR1400 conforms with this SRP. 15.0.2 Consequence Analyses for Advanced Light Water Reactors 1.9-67 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (28 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 15.1.1-15.1.4 - Decrease in Feedwater Temperature, Rev. 2 The APR1400 conforms with this SRP. 15.1.1 ~ 15.1.4 Increase in Feedwater Flow, Increase in Steam Flow, 03/2007 and Inadvertent Opening of a Steam Generator Relief or Safety Valve 15.1.5 - Steam System Piping Failures Inside and Rev. 3 The APR1400 conforms with this SRP. 15.1.5 Outside of Containment (PWR) 03/2007 15.1.5.A - Radiological Consequences of Main Steam Rev. 2 Not applicable. 15.0.3 Line Failures Outside Containment of a PWR 07/1981 SRP 15.0.3, Design Basis Accident Radiological Consequence Analyses for Advanced Light Water Reactors is applied instead of SRP 15.1.5.A.

15.2.1-15.2.5 - Loss of External Load; Turbine Trip; Rev. 2 The APR1400 conforms with this SRP. 15.2.1 ~ 15.2.5 Loss of Condenser Vacuum; Closure of Main Steam 03/2007 Isolation Valve (BWR); and Steam Pressure Regulator Failure (Closed) 15.2.6 - Loss of Nonemergency AC Power to the Rev. 2 The APR1400 conforms with this SRP. 15.2.6 Station Auxiliaries 03/2007 15.2.7 - Loss of Normal Feedwater Flow Rev. 2 The APR1400 conforms with this SRP. 15.2.7 03/2007 15.2.8 - Feedwater System Pipe Break Inside and Rev. 2 The APR1400 conforms with this SRP. 15.2.8 Outside Containment (PWR) 03/2007 15.3.1-15.3.2 - Loss of Forced Reactor Coolant Flow Rev. 2 The APR1400 conforms with this SRP. 15.3.1, 15.3.2 Including Trip of Pump Motor and Flow Controller 03/2007 Malfunctions 15.3.3-15.3.4 - Reactor Coolant Pump Rotor Seizure Rev. 3 The APR1400 conforms with this SRP. 15.3.3, 15.3.4 and Reactor Coolant Pump Shaft Break 03/2007 15.4.1 - Uncontrolled Control Rod Assembly Rev. 3 The APR1400 conforms with this SRP. 15.4.1 Withdrawal from a Subcritical or Low Power Startup 03/2007 Condition 1.9-68 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (29 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 15.4.2 - Uncontrolled Control Rod Assembly Rev. 3 The APR1400 conforms with this SRP. 15.4.2 Withdrawal at Power 03/2007 15.4.3 - Control Rod Misoperation (System Rev. 3 The APR1400 conforms with this SRP. 15.4.3 Malfunction or Operator Error) 03/2007 15.4.4-15.4.5 - Startup of an Inactive Loop or Rev. 2 The APR1400 conforms with this SRP. 15.4.4, 15.4.5 Recirculation Loop at an Incorrect Temperature, and 03/2007 Flow Controller Malfunction causing an Increase in BWR Core Flow Rate 15.4.6 - Inadvertent Decrease in Boron Concentration Rev. 2 The APR1400 conforms with this SRP. 15.4.6 in the Reactor Coolant System (PWR) 03/2007 15.4.7 - Inadvertent Loading and Operation of a Fuel Rev. 2 The APR1400 conforms with this SRP. 15.4.7 Assembly in an Improper Position 03/2007 15.4.8 - Spectrum of Rod Ejection Accidents (PWR) Rev. 3 The APR1400 conforms with this SRP. 15.4.8 03/2007 15.4.8.A - Radiological Consequences of a Control Rev. 1 Not applicable. 15.0.3 Rod Ejection Accident (PWR) 07/1981 SRP 15.0.3, Design Basis Accident Radiological Consequence Analyses for Advanced Light Water Reactors, is applied instead of SRP 15.4.8.A.

15.4.9 - Spectrum of Rod Drop Accidents (BWR) Rev. 3 Not applicable (BWR) N/A 03/2007 15.4.9.A - Radiological Consequences of Control Rod Rev. 2 Not applicable (BWR) N/A Drop Accident (BWR) 07/1981 15.5.1-15.5.2 - Inadvertent Operation of ECCS and Rev. 2 The APR1400 conforms with this SRP. 15.5.1, 15.5.2 Chemical and Volume Control System Malfunction that 03/2007 Increases Reactor Coolant Inventory 1.9-69 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (30 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 15.6.1 - Inadvertent Opening of a PWR Pressurizer Rev. 2 Not applicable. 15.6.1 Pressure Relief Valve or a BWR Pressure Relief Valve 03/2007 This SRP applies to plants adopting Pressurizer Pressure Relief Valve instead of Pressurizer Pilot Operated Safety Relief Valve.

15.6.2 - Radiological Consequences of the Failure of Rev. 2 Not applicable. 15.0.3 Small Lines Carrying Primary Coolant Outside 07/1981 SRP 15.0.3, Design Basis Accident Radiological Containment Consequence Analyses for Advanced Light Water Reactors, is applied instead of SRP 15.6.2.

15.6.3 - Radiological Consequences of Steam Rev. 2 Not applicable. 15.0.3 Generator Tube Failure 07/1981 SRP 15.0.3, Design Basis Accident Radiological Consequence Analyses for Advanced Light Water Reactors, is applied instead of SRP 15.6.3.

15.6.4 - Radiological Consequences of Main Steam Rev. 2 Not applicable (BWR) N/A Line Failure Outside Containment (BWR) 07/1981 15.6.5 - Loss-of-Coolant Accidents Resulting From Rev. 3 The APR1400 conforms with this SRP 15.6.5 Spectrum of Postulated Piping Breaks Within the 03/2007 Reactor Coolant Pressure Boundary 15.6.5.A - Radiological Consequences of a Design Rev. 1 Not applicable. 15.0.3 Basis Loss-of-Coolant Accident Including Containment 07/1981 SRP 15.0.3, Design Basis Accident Radiological Leakage Contribution Consequence Analyses for Advanced Light Water Reactors, is applied instead of SRP 15.6.5.A.

15.6.5.B - Radiological Consequences of a Design Rev. 1 Not applicable. 15.0.3 Basis Loss-of-Coolant Accident: Leakage From 07/1981 SRP 15.0.3, Design Basis Accident Radiological Engineered Safety Feature Components Outside Consequence Analyses for Advanced Light Water Containment Reactors, is applied instead of SRP 15.6.5.B.

15.6.5.D - Radiological Consequences of a Design Rev. 1 Not applicable (BWR) N/A Basis Loss-of-Coolant Accident: Leakage From Main 07/1981 Steam Isolation Valve Leakage Control System (BWR) 1.9-70 Rev. 2

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 15.7.3 - Postulated Radioactive Releases Due to Rev. 2 Not applicable. 15.0.3 Liquid-Containing Tank Failures 07/1981 SRP 15.0.3, Design Basis Accident Radiological Consequence Analyses for Advanced Light Water Reactors, is applied instead of SRP 15.7.3.

15.7.4 - Radiological Consequences of Fuel Handling Rev. 1 Not applicable. 15.0.3 Accidents 07/1981 SRP 15.0.3, Design Basis Accident Radiological Consequence Analyses for Advanced Light Water Reactors, is applied instead of SRP 15.7.4.

15.7.5 - Spent Fuel Cask Drop Accidents Rev. 2 Not applicable. 15.0.3 07/1981 SRP 15.0.3, Design Basis Accident Radiological Consequence Analyses for Advanced Light Water Reactors, is applied instead of SRP 15.7.5.

15.8 - Anticipated Transients Without Scram Rev. 2 The APR1400 conforms with this SRP. 15.8 03/2007 15.9 - Boiling Water Reactor Stability 03/2007 Not applicable (BWR) N/A 16.0 - Technical Specifications Rev. 3 The APR1400 conforms with this SRP. 16.1, 16.2, 16.3, 03/2010 16.4, 16.5 16.1 - Risk-Informed Decision Making: Technical Rev. 1 N/A. The APR1400 does not apply Risk-Information N/A Specifications 03/2007 Technical Specifications.

17.1 - Quality Assurance During the Design and Rev. 2 The APR1400 conforms with this SRP with the following 17.1, 17.5 Construction Phases 07/1981 exceptions: 17.1.8, 17.1.9, 17.1.13, and 17.1.14 are not applied in the DC phase.

17.2 - Quality Assurance During the Operations Phase Rev. 2 Not applicable. N/A 07/1981 The COL applicant is responsible for conforming with this SRP.

17.3 - Quality Assurance Program Description 08/1990 The APR1400 conforms with this SRP with exceptions. 17.3 B-6, 7, 10, and 11 are not applied in DC phase.

1.9-71 Rev. 2

APR1400 DCD TIER 2 Table 1.9-2 (32 of 32)

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SRP Section/Title Issue Date Conformance or Summary Description of Deviation DCD Tier 2 Section 17.4 - Reliability Assurance Program (RAP) Rev.1 The APR1400 conforms with this SRP. 17.4 05/2014 17.5 - Quality Assurance Program Description-Design 03/2007 The APR1400 conforms with this SRP. 17.5 Certification, Early Site Permit and New License Applicants 17.6 - Maintenance Rule Rev. 1 Not applicable (COL) N/A 08/2007 18.0 - Human Factors Engineering Rev. 2 The APR1400 conforms with this SRP. Ch. 18 03/2007 18-A -Guidance for Crediting Manual Operator 04/2014 The APR1400 conforms with this SRP. 18.6 Actions in Diversity and Defense-in-Depth (D3)

Analyses 19.0 - Probabilistic Risk Assessment and Severe Rev. 2 The APR1400 conforms with this SRP with exceptions. 19.1, 19.2 Accident Evaluation for New Reactors 06/2007 Note: SRP Acceptance Criteria for AP600 are out of the APR1400 scope.

19.1 - Determining the Technical Adequacy of Rev. 3 Only applicable to the License Amendment Requests N/A Probabilistic Risk Assessment for Risk-Informed 09/2012 after Initial Fuel Load.

License Amendment Requests after Initial Fuel Load 19.2 - Review of Risk Information Used to Support 06/2007 Not applicable. This SRP section was written to address N/A Permanent Plant-Specific Changes to the Licensing PRAs performed in support of changes proposed for Basis: General Guidance existing, already-licensed plants.

19.4 - Strategies and Guidance to Address Loss-of- 06/2014 The APR1400 conforms with this SRP. 19.4 Large Areas of the Plant Due to Explosions and Fires 19.5 - Adequacy of Design Features and Functional 04/2013 The APR1400 conforms with this SRP. 19.5 Capabilities Identified and Described for Withstanding Aircraft Impacts 1.9-72 Rev. 2

APR1400 DCD TIER 2 Table 1.9-3 (1 of 4)

APR1400 Conformance with Generic Issues (NUREG-0933)

Issue No. Title Discussion DCD Tier 2 Section 89 Stiff Pipe Clamps The stiff pipe clamps described in the generic issue, which are preloaded to N/A prevent themselves from lifting of the piping under dynamic loading conditions, are not used for the APR1400 piping design.

186 Potential Risk and Consequences For the APR1400, design for the containment polar and refueling cranes, spent 9.1.5 of Heavy Load Drops in Nuclear fuel handling crane, and auxiliary building crane preclude the dropping of heavy Power Plants loads. A critical load is defined in ASME NOG-1-2004 as any lifted load whose uncontrolled movement or release could adversely affect a nuclear safety-related (SC-1) SSC in terms of its ability to perform a required safety function, or when uncontrolled movement or release could result in potential offsite exposure in excess of 10 CFR limits.

Cranes that may be used to handle critical loads over SC-I SSCs are classified as Type I cranes as defined in ASME NOG-1-2004 and conform with the applicable requirements of that standard as well as the Crane Manufacturers Association of America (CMAA) Specification No. 70-00. Type I cranes are designed to remain in place and support the critical load during and after, a seismic event, and are equipped with single failure-proof features in conformance with the requirements of ASME NOG-1-2004, to prevent load drops.

The APR1400 cranes that do not handle critical loads over SC-I SSCs are not required to have single failure-proof features; however, any such cranes that may travel over SC-I SSCs are designed to remain in place during a seismic event.

1.9-73 Rev. 2

APR1400 DCD TIER 2 Table 1.9-3 (2 of 4)

Issue No. Title Discussion DCD Tier 2 Section 186 Potential Risk and Consequences Cranes that handle critical loads as well as non-critical loads conform with the of Heavy Load Drops in Nuclear applicable requirements of ASME NOG-1-2004 and CMAA Specification No.

Power Plants (cont.) 70-00 or CMAA Specification No. 74-04 for their applicable lifts. Further, cranes are designed according to the crane structural standard and so structured as to prevent diversion and derailment. In addition, in the measures against earthquake, drop prevention design is employed based on earthquake design criteria.

Therefore, load drops and derailment of cranes do not represent credible sources of missiles that would jeopardize safety-related SSCs, and load drop missiles are not postulated. The significance of crane operation and restricted load movement around the reactor vessel is stressed to those involved with heavy load lifts. Anticipated heavy load movements are analyzed as required by NUREG-0612 and safe load paths defined. However, all specific loads and load paths cannot be defined prior to the operations. For these cases, it is anticipated that safe load path considerations are based on comparison with analyzed cases, previously defined safe movement areas, and previously defined restricted areas and reviewed by the COL applicants plant review board.

Load handling procedures - Movements of heavy loads are controlled to protect safety-related SSCs. Load handling operations for heavy loads that are or could be handled over or in proximity to irradiated fuel or safe shutdown equipment will be controlled by written procedures.

1.9-74 Rev. 2

APR1400 DCD TIER 2 Table 1.9-3 (3 of 4)

Issue No. Title Discussion DCD Tier 2 Section 186 Potential Risk and Consequences As a minimum, procedures are used for handling loads with spent fuel cask of Heavy Load Drops in Nuclear bridge crane and polar crane, and for the loads listed in Table 3-1 of NUREG-Power Plants (cont.) 0612. It is anticipated that each procedure will address the following:

• Equipment required to handle load (e.g., special lifting device, slings, shackles, turnbuckles, clevises, load cell)

• Requirements for crane operator and rigger qualification

• Requirements for inspection prior to load movement and acceptance criteria for inspection

• Defined safe load path and provisions to provide visual reference to the crane operator and/or signal person of the safe load path envelope

• Specific steps and proper sequence to be followed for handling load

• Precautions, limitations, prerequisites, and/or initial conditions associated with movement of the load

• Slings and other equipment used to make a complete lifting device specified in the load handling procedures, which conform with NUREG-0612

• Equipment layout drawings showing the safe load path, which are used to define safe load paths in load handling procedures; deviations from defined safe load paths require a written alternative procedure approved by the COL applicants plant review board These considerations and commitments, as well as the other design and operational material presented above, are intended to prevent the types of events that are the subject of this generic issue. The APR1400 design is attentive to any new NRC operational guidance related to this issue.

1.9-75 Rev. 2

APR1400 DCD TIER 2 Table 1.9-3 (4 of 4)

Issue No. Title Discussion DCD Tier 2 Section 189 Susceptibility of Ice Condenser Not applicable (applicable to ice condenser containment only) N/A Containments to Early Failure from Hydrogen Combustion During Severe Accident 191 Assessment of Debris The APR1400 is designed in accordance with NRC RG 1.82, Revision 4, the 6.8.4.5, Accumulation on PWR Sump methodology of NEI 04-07, and the NRCs Safety Evaluation Report (SER) for Performance NEI 04-07. Four redundant passive-type strainers are installed in the IRWST, which has a broad footprint for sufficient surface area. Insulation and coating debris is estimated by the NEI 04-07 methodology, and 200 pounds of latent debris is assumed to reach each strainer location. Trisodium phosphate (TSP) is selected as the agent for pH control in the recirculation water inside the holdup volume tank (HVT), to mitigate the chemical effect that might be caused during long-term cooling.

193 BWR ECCS Suction Concerns Not applicable (BWR) N/A 199 Implications of Updated Not applicable (COL) 2.5 Probabilistic seismic Hazard Estimates in Central and Eastern United States 1.9-76 Rev. 2

APR1400 DCD TIER 2 Table 1.9-4 (1 of 10)

APR1400 Conformance with Additional TMI-Related Requirements (10 CFR 50.34(f))

10 CFR 50.34(f)

Item / Issue No. Requirements Conformance Discussion DCD Tier 2 Section (1)(i) / II.B.8 Perform a plant/site specific probabilistic risk assessment, the aim of The APR1400 conforms with this 19.1 which is to seek such improvements in the reliability of core and TMI-related requirement.

containment heat removal systems as are significant and practical and do not impact excessively on the plant.

(1)(ii) / II.E.1.1 Perform an evaluation of the proposed auxiliary feedwater system Table 15.0-11, (AFWS), to include (PWRs only): 19.1

a. A simplified AFWS reliability analysis using event-tree and a. The APR1400 conforms with this fault-tree logic techniques TMI-related requirement N/A

b. A design review of AFWS b. Not applicable N/A

c. An evaluation of AFWS flow design bases and criteria c. Not applicable 1.9-77 Rev. 2

APR1400 DCD TIER 2 Table 1.9-4 (2 of 10) 10 CFR 50.34(f)

Item / Issue No. Requirements Conformance Discussion DCD Tier 2 Section (1)(iii) / Perform an evaluation of the potential for and impact of reactor RCP seal integrity can be maintained 9.2.2.2.4.5, II.K.2.16, coolant pump seal damage following small-break LOCA with loss of by either of two independent sources Table 15.0-11 II.K.3.25 offsite power. If damage cannot be precluded, provide an analysis of cooling water: the seal injection of the limiting small-break loss-of-coolant accident with subsequent flow from the chemical and volume reactor coolant pump seal damage. control system (CVCS) or component cooling water (CCW).

In the event of a loss of offsite AC power, the seal injection can be restored by aligning the emergency diesel generators (EDGs) power to the charging pumps or auxiliary charging pump (ACP). The CCW pumps restart in accordance with the EDG load sequencing to provide seal cooling.

During a complete loss of AC power (loss of offsite power with loss of the EDGs), power can be supplied to the ACP from the onsite AAC power source to provide the RCP seal injection. Therefore, The APR1400 conforms with this TMI-related requirement.

(1)(iv) / II.K.3.2 Perform an analysis of the probability of a small-break loss-of- Not applicable. N/A coolant accident (LOCA) caused by a stuck-open power-operated There is no PORV for the APR1400.

relief valve (PORV). If this probability is a significant contributor to the probability of small-break LOCAs from all causes, provide a description and evaluation of the effect on small-break LOCA probability of an automatic PORV isolation system that would operate when the reactor coolant system pressure falls after the PORV has opened.

1.9-78 Rev. 2

APR1400 DCD TIER 2 Table 1.9-4 (3 of 10) 10 CFR 50.34(f)

Item / Issue No. Requirements Conformance Discussion DCD Tier 2 Section (2)(i) / I.A.4.2 Provide simulator capability that correctly models the control room Not applicable (COL) N/A and includes the capability to simulate small-break LOCAs.

(2)(ii) / I.C.9 Establish a program, to begin during construction and follow into Not applicable (COL) N/A operation, for integrating and expanding current efforts to improve plant procedures. The scope of the program shall include emergency procedures, reliability analyses, human factors engineering, crisis management, operator training, and coordination with Institute of Nuclear Power Operations (INPO) and other industry efforts.

(2)(iii) / I.D.1 Provide, for Commission review, a control room design that reflects The APR1400 conforms with this 18.7 state-of-the-art human factors principles prior to committing to TMI-related requirement.

fabrication or revision of fabricated control room panels and layouts.

(2)(iv) / I.D.2 Provide a plant safety parameter display console that displays to The APR1400 conforms with this 7.5.2.5 operators a minimum set of parameters defining the safety status of TMI-related requirement. 18.7 the plant, capable of displaying a full range of important plant parameters and data trends on demand, and capable of indicating when process limits are being approached or exceeded.

(2)(v) / I.D.3 Provide for automatic indication of the bypassed and operable status The APR1400 conforms with this 7.1.2.5, 7.5.2.3, of safety systems. TMI-related requirement. 7.6.2.1, 8.3.1.1.3, 8.3.1.2.3, 8.3.2.2.3 (2)(vi) / II.B.1 Provide the capability of high-point venting of noncondensable gases The high-point vent system is installed 5.4.12.1 from the RCS, and other systems that may be required to maintain to meet this requirement.

adequate core cooling. Systems to achieve this capability shall be capable of being operated from the control room, and their operation shall not lead to an unacceptable increase in the probability of LOCA or an unacceptable challenge to containment integrity.

1.9-79 Rev. 2

APR1400 DCD TIER 2 Table 1.9-4 (4 of 10) 10 CFR 50.34(f)

Item / Issue No. Requirements Conformance Discussion DCD Tier 2 Section (2)(vii) / II.B.2 Perform radiation and shielding design reviews of spaces around The APR1400 conforms with this 12.4.1.2.7 systems that may, as a result of an accident, contain accident source TMI-related requirement.

term11 radioactive materials, and design as necessary to permit adequate access to important areas and to protect safety equipment from the radiation environment.

(2)(viii) / II.B.3 Provide a capability to promptly obtain and analyze samples from the The APR1400 has a post-accident 9.3.2.1, RCS and containment that may contain accident source term11 sampling system to conform with this Table 15.0-11 radioactive materials without radiation exposures to any individual action item.

exceeding 5 rems to the whole body or 50 rems to the extremities.

Materials to be analyzed and quantified include certain radionuclides that are indicators of the degree of core damage (e.g., noble gases, radioiodines and cesiums, nonvolatile isotopes), hydrogen in the containment atmosphere, dissolved gases, chloride, and boron concentrations.

1.9-80 Rev. 2

APR1400 DCD TIER 2 Table 1.9-4 (5 of 10) 10 CFR 50.34(f)

Item / Issue No. Requirements Conformance Discussion DCD Tier 2 Section (2)(ix) / II.B.8 Provide a system for hydrogen control that can safely accommodate The APR1400 conforms with this 6.2.5.1 hydrogen generated by the equivalent of a 100% fuel-clad metal TMI-related requirement.

water reaction. Preliminary design information on the tentatively preferred system option of those being evaluated in paragraph (f)(1)(xii) of this section is sufficient at the construction permit stage.

The hydrogen control system and associated systems shall provide, with reasonable assurance, that:

a. Uniformly distributed hydrogen concentrations in the containment do not exceed 10% during and following an accident that releases an equivalent amount of hydrogen as would be generated from a 100% fuel clad metal-water reaction, or that the post-accident atmosphere will not support hydrogen combustion.

b. Combustible concentrations of hydrogen will not collect in areas where unintended combustion or detonation could cause loss of containment integrity or loss of appropriate mitigating features.

c. Equipment necessary for achieving and maintaining safe shutdown of the plant and maintaining containment integrity will perform its safety function during and after being exposed to the environmental conditions attendant with the release of hydrogen generated by the equivalent of a 100% fuel-clad metal water reaction including the environmental conditions created by activation of the hydrogen control system.

d. If the method chosen for hydrogen control is a post-accident inerting system, inadvertent actuation of the system can be safely accommodated during plant operation.

1.9-81 Rev. 2

APR1400 DCD TIER 2 Table 1.9-4 (6 of 10) 10 CFR 50.34(f)

Item / Issue No. Requirements Conformance Discussion DCD Tier 2 Section (2)(x) / II.D.1 Provide a test program and associated model development and Performance testing for the POSRV is 5.2.2.10 conduct tests to qualify reactor coolant system relief and safety valves performed on both normal and and, for PWRs, PORV block valves, for all fluid conditions expected accident conditions, excluding under operating conditions, transients and accidents. Consideration anticipated transient without scram of anticipated transients without scram (ATWS) conditions shall be (ATWS), to provide the stable valve included in the test program. Actual testing under ATWS conditions operation.

need not be carried out until subsequent phases of the test program are developed.

(2)(xi) / II.D.3 Provide direct indication of relief and safety valve position (open or The APR1400 conforms with this 5.2.2.1.1, closed) in the control room. TMI-related requirement. 5.2.2.8, 7.1.2.6 (2)(xii) / II.E.1.2 Provide automatic and manual auxiliary feedwater (AFW) system The APR1400 conforms with this 7.1.2.7, initiation, and provide auxiliary feedwater system flow indication in TMI-related requirement. 7.2, the control room. (PWRs only) Table 15.0-11 (2)(xiii) / Provide pressurizer heater power supply and associated motive and The APR1400 conforms with this 8.3.1.1.2, 8.3.1.2.3 II.E.3.1 control power interfaces sufficient to establish and maintain natural TMI-related requirement.

circulation in hot standby conditions with only onsite power available. (PWRs only)

(2)(xiv) / Provide containment isolation systems that: The APR1400 conforms with this 6.2.4.2, II.E.4.2 a. Ensure all non-essential systems are isolated automatically by TMI-related requirement. 7.1.2.8, the containment isolation system 7.2, 7.5

b. For each non-essential penetration (except instrument)

c. Do not result in reopening of the containment isolation

d. Utilize a containment setpoint pressure for initiating containment isolation as low as is compatible with normal

e. Include automatic closing on a high radiation signal for all systems that provide a path to the environs 1.9-82 Rev. 2

APR1400 DCD TIER 2 Table 1.9-4 (7 of 10) 10 CFR 50.34(f)

Item / Issue No. Requirements Conformance Discussion DCD Tier 2 Section (2)(xv) / II.E.4.4 Provide a capability for containment purging/venting designed to As described in NUREG-0933, under 9.4.6.2 minimize the purging time consistent with as low as reasonably item II.E.4.4 (4), this item required achievable (ALARA) principles for occupational exposure. Provide the U.S. NRC to generically evaluate and demonstrate high assurance that the purge system will reliably the radiological consequences of isolate under accident conditions. containment purging of nuclear plants while in the power operation mode.

It was envisioned that, as a result of this evaluation, new requirements would be needed beyond those in SRP 6.2.4 and BTP 6-4. The NRC subsequently determined that this issue was a low-priority item; it was then resolved without issuance of new requirements. The valve operability guidance provided in SRP Section 6.2.4 and BTP 6-4, Rev.3, dated March 2007, was considered adequate by the U.S. NRC.

The APR1400 conforms with SRP 6.2.4 and BTP 6-4 (Rev. 3, 03/2007).

(2)(xvii) / II.F.1 Provide instrumentation to measure, record, and readout in the control The APR1400 conforms with this 7.1.2.9, room (A) containment pressure, (B) containment water level, (C) TMI-related requirement. 11.5.1.2, containment hydrogen concentration, (D) containment radiation 12.3.4.1.5, intensity (high level), and (E) noble gas effluents at all potential, Table 15.0-11 accident release points. Provide for continuous sampling of radioactive iodines and particulates in gaseous effluents from all potential accident release points, and for onsite capability to analyze and measure these samples.

1.9-83 Rev. 2

APR1400 DCD TIER 2 Table 1.9-4 (8 of 10) 10 CFR 50.34(f) DCD Tier 2 Item / Issue No. Requirements Conformance Discussion Section (2)(xviii) / II.F.2 Provide instruments that provide in the control room an unambiguous The APR1400 conforms with this 7.1.1.5, indication of inadequate core cooling, such as primary coolant TMI-related requirement based on 7.1.2.10, saturation meters in PWRs, and a suitable combination of signals NUREG-0737. 7.5.1.2, from indicators of coolant level in the reactor vessel and in-core Table 15.0-11 thermocouples in PWRs and BWRs.

(2)(xix) / II.F.3 Provide instrumentation adequate for monitoring plant conditions The APR1400 conforms with this 7.1.2.11 following an accident that includes core damage. TMI-related requirement based on Table 15.0-11 NRC RG 1.97 Rev. 4.

(2)(xx) / II.G.1 Provide power supplies for pressurizer relief valves, block valves, The APR1400 conforms with this 7.1.2.12, and level indicators such that (A) level indicators are powered from TMI-related requirement. 7.4.2, vital buses; (B) motive and control power connections to the 8.3.1.2.3 emergency power sources are through devices qualified in accordance with requirements applicable to systems important to safety, and (C) electric power is provided from emergency power sources. (PWRs only)

(2)(xxv) / Provide an onsite Technical Support Center, an onsite Operational The APR1400 conforms with this 9.5.2.1 III.A.1.2 Support Center, and for construction permit applications only, a near- TMI-related requirement.

site Emergency Operations Facility.

(2)(xxvi) / Provide for leakage control and detection in the design of systems The APR1400 conforms with this 9.3.3 III.D.1.1 outside containment that contain (or might contain) accident source TMI-related requirement.

term radioactive materials following an accident. Applicants shall submit a leakage control program, including an initial test program, a schedule for re-testing these systems, and the actions to be taken for minimizing leakage from such systems. The goal is to minimize potential exposures to workers and public, and to provide reasonable assurance that excessive leakage will not prevent the use of systems needed in an emergency.

1.9-84 Rev. 2

APR1400 DCD TIER 2 Table 1.9-4 (9 of 10) 10 CFR 50.34(f)

Item / Issue No. Requirements Conformance Discussion DCD Tier 2 Section (2)(xxvii) / Provide for monitoring of inplant radiation and airborne radioactivity The APR1400 conforms with this 11.5.1.2, III.D.3.3 as appropriate for a broad range of routine and accident conditions. TMI-related requirement. 12.3.4 (2)(xxviii) / Evaluate potential pathways for radioactivity and radiation that may The APR1400 conforms with this 6.4.2.5, III.D.3.4 lead to control room habitability problems under accident conditions TMI-related requirement. 15.6.5.5 resulting in an accident source term release, and make necessary design provisions to preclude such problems.

(3)(i) / I.C.5 Provide administrative procedures for evaluating operating, design The APR1400 conforms with this 13.5 and construction experience and for ensuring that applicable TMI-related requirement.

important industry experiences will be provided in a timely manner to those designing and constructing the plant.

(3)(ii) / I.F.1 Ensure that the quality assurance (QA) list required by Criterion II, Not applicable (COL) N/A Appendix. B, 10 CFR 50 includes all structures, systems, and components important to safety.

(3)(iii) / I.F.2 Establish a quality assurance (QA) program based on consideration of The APR1400 conforms with this Ch. 17 (A) ensuring independence of the organization performing checking TMI-related requirement.

functions from the organization responsible for performing the functions, (B) performing quality assurance/quality control functions at construction sites to the maximum feasible extent, (C) including QA personnel in the documented review of and concurrence in quality related procedures associated with design, construction and installation, (D) establishing criteria for determining QA programmatic requirements, (E) establishing qualification requirements for QA and QC personnel, (F) sizing the QA staff commensurate with its duties and responsibilities, (G) establishing procedures for maintenance of as-built documentation, and (H) providing a QA role in design and analysis activities.

1.9-85 Rev. 2

APR1400 DCD TIER 2 Table 1.9-4 (10 of 10) 10 CFR 50.34(f)

Item / Issue No. Requirements Conformance Discussion DCD Tier 2 Section (3)(iv) / II.B.8 Provide one or more dedicated containment penetrations: The APR1400 conforms with this 19.2.3.3.8 Equivalent in size to a single 3-foot-diameter opening, in order not to TMI-related requirement.

preclude future installation of systems to prevent containment failure, such as a filtered vented containment system.

(3)(vi) / II.E.4.1 For plant designs with external hydrogen recombiners: The dedicated containment penetration 6.2.5 Provide redundant dedicated containment penetrations so that, is not necessary because the APR1400 assuming a single failure, the recombiner systems can be connected to has applied the passive autocatalytic the containment atmosphere. recombiners for hydrogen control, which are located inside the containment.

(3)(vii) / II.J.3.1 Provide a description of the management plan for design and Not applicable (COL) N/A construction activities, to include (A) the organizational and management structure singularly responsible for direction of design and construction of the proposed plant, (B) technical resources director by the applicant, (C) details of the interaction of design and construction within the applicants organization and the manner by which the applicant will ensure close integration of the architect engineer and the nuclear steam supply vendor, (D) proposed procedures for handling the transition to operation, (E) the degree of top level management oversight and technical control to be exercised by the applicant during design and construction, including the preparation and implementation of procedures necessary to guide the effort.

1.9-86 Rev. 2

APR1400 DCD TIER 2 Table 1.9-5 Generic Communications Applicability to APR1400 GC No. Title Comment DCD Tier 2 Section GL 2007-01 Inaccessible or Underground Power Cable Failures that Disable The APR1400 conforms with this 8.3.1.1.10 Accident Mitigation Systems or Cause Plant Transients Generic Letter GL 2008-01 Managing Gas Accumulation in Emergency Core Cooling, Decay The APR1400 conforms with this 6.2.2, 6.3.2.5.2 Heat Removal, and Containment Spray Systems. Generic Letter.

BL 2007-01 Security Officer Attentiveness Not applicable (COL) N/A BL 2011-01 Mitigation Strategies Not applicable N/A 1.9-87 Rev. 2

APR1400 DCD TIER 2 Table 1.9-6 (1 of 2)

Summary of SECY Documents Provided in Section C.I.1.9.5 of NRC RG 1.206 SECY Paper No. Title Discussion 89-013 Design Requirements Related to the Evolutionary Advanced Light-Water See Table 1.9-7 for SECY-93-087 Reactors (ALWRs)90-016 Evolutionary Light-Water Reactor (ELWR) Certification Issues and Their See Table 1.9-7 for SECY-93-087 Relationship to Current Regulatory Requirements.90-241 Level of Detail Required for Design Certification under 10 CFR 52. The recommendations in this SECY were incorporated into 10 CFR Part 52. Conformance is addressed in Section 1.9 of this DCD Tier 2.90-377 Requirements for Design Certification under 10 CFR 52. The recommendations in this SECY were incorporated into 10 CFR Part 52. Conformance is addressed in Section 1.9 of this DCD Tier 2.91-074 Prototype Decisions for Advanced Reactor Designs. APR1400 does not contain Significant deviation from reference plants of standard technologies, and thus does not require prototypical demonstration.91-178 ITAAC for Design Certifications and Combined Licenses. The recommendations in this SECY were incorporated into 10 CFR Part 52. Conformance is addressed in Section 1.9.1 through 1.9.5 of this DCD Tier 2.91-210 ITAAC Requirements for Design Review and Issuance of FDA. The recommendations in this SECY were incorporated into 10 CFR Part 52. Conformance is addressed in Section 1.9 of this DCD Tier 2.91-229 Severe Accident Mitigation Design Alternatives for Certified Standard Severe accidents are addressed in Section 19 of this DCD Designs. Tier 2, and severe accident mitigation design alternatives (SAMDAs) are addressed in Subsection 19.2.6.

1.9-88 Rev. 2

APR1400 DCD TIER 2 Table 1.9-6 (2 of 2)

SECY Paper No. Title Discussion 92-053 Use of Design Acceptance Criteria During the 10 CFR 52 Design The recommendations in this SECY were incorporated into Certification Reviews. 10 CFR Part 52. Conformance is addressed in Subsection 1.9.1 through 1.9.5 of this DCD Tier 2.92-092 The Containment Performance Goal, External Events Sequences, and the This SECY does not impose any new requirements.

Definition of Containment Failure for Advanced LWRs.93-087 Policy, Technical, and Licensing Issues Pertaining to Evolutionary and See Table 1.9-7.

Advanced Light-Water Reactor (ALWR) Designs.94-084 Policy and Technical Issues Associated with the Regulatory Treatment of Not applicable Non-Safety Systems (RTNSS) in Passive Plant Design.94-302 Source-Term-Related Technical and Licensing Issues Relating to The APR1400 conforms with the positions presented in the Evolutionary and Passive Light-Water-Reactor Designs. SECY. the significant issues will be addressed in applicable to the APR1400 DCD Tier 2.95-132 Policy and Technical Issues Associated with Regulatory Treatment of Not applicable (AP1000 design only).

Non-Safety Systems in Passive Plant Designs.

1.9-89 Rev. 2

APR1400 DCD TIER 2 Table 1.9-7 (1 of 4)

Conformance with SECY-93-087 Item No. Title Discussion I.A Use of a Physically Based Source Term Addressed for the APR1400 in DCD Tier 2 Subsection 3.11.6 and Appendix 15A.

I.B Anticipated Transient Without Scram Addressed for the APR1400 in DCD Tier 2 Section 15.8.

I.C Mid-Loop Operation Addressed for the APR1400 in DCD Tier 2 Subsection 5.4.7.2.6 and 19.2.2.2.

I.D Station Blackout Addressed for the APR1400 in DCD Tier 2 Section 8.4.

I.E Fire Protection Addressed for the APR1400 in DCD Tier 2 Subsections 3.1.1 and 9.5.1.

I.F Intersystem Loss-of-Coolant Accident Addressed for the APR1400 in DCD Tier 2 Subsections 5.2.5.4 and 19.2.2.5.

I.G Hydrogen Control Addressed for the APR1400 in DCD Tier 2 Subsections 6.2.5 and 19.2.3.3.2.

Non-safety-related HMS consisting of igniters and passive autocatalytic recombiners are located in containment adequately.

I.H Core Debris Coolability Addressed for the APR1400 in DCD Tier 2 Subsection 19.2.3.3.3.

Core coolability is confirmed using MAAP code.

I.I High-Pressure Core Melt Ejection Addressed for the APR1400 in DCD Tier 2 Subsection 19.2.3.3.4.

1.9-90 Rev. 2

APR1400 DCD TIER 2 Table 1.9-7 (2 of 4)

Item No. Title Discussion I.J Containment Performance Designed robustly the APR1400 containment to withstand containment pressure challenges.

I.K Dedicated Containment Vent Penetration Dedicated containment vent is not provided.

I.L Equipment Survivability Addressed for the APR1400 in DCD Tier 2, Section 19.2.

I.M Elimination of Operating-Basis Earthquake Addressed for the APR1400 in DCD Tier 2, Subsection 3.2.6.

I.N Inservice Testing of Pumps and Valves Addressed for the APR1400 in DCD Tier 2, Subsections 3.1.4 and 3.9.6, Sections 6.6 and 13.4, and Chapter 16.

II.A Industry Codes and Standards Addressed for the APR1400 in DCD Tier 2, Subsection 3.2.8.

II.B Electrical Distribution The APR1400 conforms with the requirement of SECY 91-078.

II.C Seismic Hazard Curves and Design Parameters Not applicable (information only)

II.D Leak-Before-Break Addressed for the APR1400 in DCD Tier 2, Subsection 3.6.3.

II.E Classification of Main Steamlines in Boiling Water Reactors Not applicable (BWR)

II.F Tornado Design Basis The design basis tornado with a maximum wind speed of 230 mph is employed in accordance with Revision 1 of NRC RG 1.76.

1.9-91 Rev. 2

APR1400 DCD TIER 2 Table 1.9-7 (3 of 4)

Item No. Title Discussion II.G Containment Bypass Addressed for the APR1400 in DCD Tier 2, Subsection 19.2.2.5.

The principal contributors of containment bypass are steam generator tube ruptures (SGTRs) with MSSVs or ADVs and interfacing-system LOCAs (ISLOCAs). The APR1400 is designed to prevent and mitigate the following accidents:

• The APR1400 has performed the Interfacing System LOCA (ISLOCA) evaluation to address the issue of containment bypass.

Improvements made to the APR1400 resulting from the ISLOCA evaluation include:

• Increasing the design pressure rating of equipment or systems to at least 900 psig.

• Adding equipment and instrumentation that alert the operator to an ISLOCA challenge, or terminate and limit the scope of an ISLOCA.

• Improvement the capability for leak testing pressure isolation valves.

• Pressure isolation valve position indication and control in the control room.

• High pressure alarms to warn the operator when rising pressure approaches the design pressure of low-pressure systems.

All the improvements made to the APR1400 as a result of the ISLOCA evaluation are addressed in this DCD, Tier 2, Subsection 19.2.2.5.

II.H Containment Leak Rate Testing Not applicable. Containment Leakage Rate Testing conforms to NRC RG 1.163 (Refer to Subsection 6.2.6 and TS 5.5.16.)

II.I Post-Accident Sampling System Conformance is described in Subsection 9.3.2.

II.J Level of Detail The APR1400 has the level of detail of information required to acquire design certification.

II.K Prototyping Not applicable (information only) 1.9-92 Rev. 2

APR1400 DCD TIER 2 Table 1.9-7 (4 of 4)

Item No. Title Discussion II.L ITAAC Development guidance for ITAAC is addressed in Section 14.3 and ITAAC for each system is described in Tier 1 of this DCD.

II.M Reliability Assurance Program The APR1400 reliability assurance program, addressing the requirements appropriate for design certification is presented in the DCD Tier 2, Section 17.4.

II.N Site-Specific Probabilistic Risk Assessments and Analysis of Addressed for the APR1400 in DCD Tier 2, Section 19.1. PRA covers, External Events internal fire events, and internal flooding events as well as internal events.

Seismic events were evaluated using a PRA-based Seismic Margin Analysis (SMA) rather than a seismic PRA. The COL applicant is to perform site-specific PRA evaluations to address any site-specific hazards.

II.O Severe Accident Mitigation Design Alternatives Addressed for the APR1400 in DCD Tier 2, Subsection 19.2.6.

II.P Generic Rulemaking Related to Design Certification Not applicable (information only)

II.Q Defense Against Common-Mode Failures in Digital Addressed for the APR1400 in DCD Tier 2, Subsection 7.1.2.36, Instrumentation and Control Systems Table 7.1-1, and Subsections 7.3.2.4, 7.8.2.1, 7.8.2.2, and 7.8.2.3.

II.R Steam Generator Tube Ruptures Addressed for the APR1400 in DCD Tier 2, Subsection 15.6.3.

II.S PRA Beyond Design Certification This item is not applicable to the design certification phase, since it requires site specific information. This item will be addressed in the COL phase. See Subsection 19.1.9.

II.T Control Room Annunciator (Alarm) Reliability Addressed for the APR1400 in DCD Tier 2, Subsection 7.1.2.37, Table 7.1-1, and Subsection 7.5.2.4.

III.E Control Room Habitability Not applicable III.F Radionuclide Attenuation Not applicable 1.9-93 Rev. 2

APR1400 DCD TIER 2 Table 1.9-8 (1 of 19)

APR1400 Strategies for Addressing Tier 1, 2 and 3 NTTF Recommendations NTTF Rec. NRC Recommendations/Requirements in SECY-11-0093, SECY Applicable No 0137, SECY-12-0025, SECY-12-0095, EA-12-049, EA-12-051 APR1400 Design DCD Section COL Action Note Tier 1 (Actions to be taken without delay) 2.1 Seismic Reevaluation NA NA COL 19.3(1) Request for a) Evaluate the potential impacts of the newly released Central information and Eastern United States Seismic Source Characterization via 50.54 (CEUS-SSC) model, with potential local and regional (f) letter.

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.

1.9-94 Rev. 2

APR1400 DCD TIER 2 Table 1.9-8 (2 of 19)

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

• Perform a reevaluation of all appropriate external flooding information sources, including the effects from local intense precipitation via 50.54 on the site, probable maximum flood (PMF) on stream and (f) letter.

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

• Perform seismic walkdowns in order to identify and address information plant specific degraded, non-conforming, or unanalyzed via 50.54 conditions and verify the adequacy of strategies, monitoring, (f) letter.

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.

1.9-95 Rev. 2

APR1400 DCD TIER 2 Table 1.9-8 (3 of 19)

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

• Perform flood protection walkdowns using an NRC-endorsed information walkdown methodology, via 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 Technical Report See DCD COL (NTTF Recommendations) Initiate rulemaking to revise 10 CFR APR1400-E-P-NR- Section 19.3.2.3 19.3(3),

50.63 to require each operating and new reactor licensee to (1) 14005-P, Rev 2; 19.3(4) and establish a minimum coping time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for a loss of all ac Section 5.1.2 19.3(9) 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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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 offsite transportation infrastructure associated with significant natural disasters.

1.9-96 Rev. 2

APR1400 DCD TIER 2 Table 1.9-8 (4 of 19)

NTTF Rec. NRC Recommendations/Requirements in SECY-11-0093, SECY Applicable No 0137, SECY-12-0025, SECY-12-0095, EA-12-049, EA-12-051 APR1400 Design DCD Section COL Action Note 4.2 Mitigation Strategies for Beyond-Design-Basis External Events See Technical Report See DCD COL (EA-12-049) APR1400-E-P-NR- Section 19.3.2.3 19.3(3),

14005-P, Rev 2; 19.3(4) and

1. Licensees shall develop, implement and maintain guidance and Section 5.1.2 19.3(5) 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.

1.9-97 Rev. 2

APR1400 DCD TIER 2 Table 1.9-8 (5 of 19)

NTTF Rec. NRC Recommendations/Requirements in SECY-11-0093, SECY Applicable No 0137, SECY-12-0025, SECY-12-0095, EA-12-049, EA-12-051 APR1400 Design DCD Section COL Action Note 5.1 Reliable Hardened Vents for Mark I and Mark II containments NA NA NA 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 SFP Instrumentation See Technical Report See DCD COL (EA-12-051 to COL Holder) APR1400-E-P-NR- Section 19.3.2.4 19.3(12) 14005-P, Rev 2; Licensee requires reliable indication of the water level in associate Section 5.1.3 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 make-up water addition should no longer be deferred.

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

1.9-98 Rev. 2

APR1400 DCD TIER 2 Table 1.9-8 (6 of 19)

NTTF Rec. NRC Recommendations/Requirements in SECY-11-0093, SECY- Applicable No 11-0137, SECY-12-0025, SECY-12-0095, EA-12-049, EA-12-051 APR1400 Design DCD Section COL Action Note 7.1 1.1 Arrangement: The spent fuel pool level instrument (cont.) 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 mobile 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. On-site 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 offsite resource availability is reasonably assured.

1.9-99 Rev. 2

APR1400 DCD TIER 2 Table 1.9-8 (7 of 19)

NTTF Rec. NRC Recommendations/Requirements in SECY-11-0093, SECY- Applicable No 11-0137, SECY-12-0025, SECY-12-0095, EA-12-049, EA-12-051 APR1400 Design DCD Section COL Action Note 7.1 1.4 Accuracy: The instrument shall maintain its designed (cont.) 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 operating procedures, NA NA COL 19.3(13) 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.

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.

1.9-100 Rev. 2

APR1400 DCD TIER 2 Table 1.9-8 (8 of 19)

NTTF Rec. NRC Recommendations/Requirements in SECY-11-0093, SECY- Applicable No 11-0137, SECY-12-0025, SECY-12-0095, EA-12-049, EA-12-051 APR1400 Design DCD Section COL Action Note 8 3. Order licensees to modify each plants technical (cont.) 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.

9.3 Emergency Preparedness NA NA COL 19.3(14)

(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 on-site communications systems and their required normal and/or backup power supplies,

• 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, 1.9-101 Rev. 2

APR1400 DCD TIER 2 Table 1.9-8 (9 of 19)

NTTF Rec. NRC Recommendations/Requirements in SECY-11-0093, SECY- Applicable No 11-0137, SECY-12-0025, SECY-12-0095, EA-12-049, EA-12-051 APR1400 Design DCD Section COL Action Note 9.3 2. Describe any interim actions that have been taken or are (cont.) 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.

Staffing NA COL 19.3(15)

1. Provide an assessment of the on-site 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 on-site and augmented staff available to implement the strategies as discussed in the emergency plan and/or described in plant operating procedures. The following functions are requested to be assessed:

• How on-site staff will move back-up equipment (e.g.,

pumps, generators) from alternate on-site 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.

1.9-102 Rev. 2

APR1400 DCD TIER 2 Table 1.9-8 (10 of 19)

NTTF Rec. NRC Recommendations/Requirements in SECY-11-0093, SECY- Applicable No 11-0137, SECY-12-0025, SECY-12-0095, EA-12-049, EA-12-051 APR1400 Design DCD Section COL Action Note 9.3

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

2. Provide an implementation schedule of the time needed to conduct the on-site 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.

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.).

1.9-103 Rev. 2

APR1400 DCD TIER 2 Table 1.9-8 (11 of 19)

NTTF Rec. NRC Recommendations/Requirements in SECY-11-0093, SECY- Applicable No 11-0137, SECY-12-0025, SECY-12-0095, EA-12-049, EA-12-051 APR1400 Design DCD Section COL Action Note Filtration of Containment Vents NA NA NA 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 NA NA COL 19.3(1)

(SECY-12-0025) and 19.3(2)

1. Include UHS systems in the reevaluation and walkdowns of site-specific seismic and flooding hazards using the methodology described in SECY-11-0137, and identify actions that have been taken, or are planned, to address plant-specific issues associated with the updated seismic and flooding hazards in conjunction with the resolution of NTTF Recommendations 2.1 and 2.3.

2. Incorporate the loss of UHS as a design assumption in the See Technical See DCD COL 19.3(3),

resolution of station blackout rulemaking activities in Report APR1400-E- Section 19.3(4) and conjunction with the resolution of NTTF Recommendation P-NR-14005-P, Rev 19.3.2.3 19.3(9) 4.1. 0; Section 5.1.2

3. Provide mitigating measures for beyond-design-basis external See Technical See DCD COL 19.3(3),

events to also include a loss of access to the normal UHS in Report APR1400-E- Section 19.3(4) and conjunction with the resolution of NTTF Recommendation P-NR-14005-P, Rev 19.3.2.3 19.3(9) 4.2. 0; Section 5.1.2 1.9-104 Rev. 2

APR1400 DCD TIER 2 Table 1.9-8 (12 of 19)

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

4. Include UHS systems in the reevaluation of site-specific NA NA Refer to Tier 2 natural external hazards, and identify actions that have been Recommendation 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 (See NA NA (SECY-12-0025) Technical Report APR1400- E-P-NR-

1. Continue stakeholder interactions to discuss the technical 14005-P, Rev 2) basis and acceptance criteria for conducting a reevaluation of site-specific 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.

1.9-105 Rev. 2

APR1400 DCD TIER 2 Table 1.9-8 (13 of 19)

NTTF Rec. NRC Recommendations/Requirements in SECY-11-0093, SECY- Applicable No 11-0137, SECY-12-0025, SECY-12-0095, EA-12-049, EA-12-051 APR1400 Design DCD Section COL Action Note 7 SFP Makeup Capability (NTTF 7.2, 7.3, 7.4, and 7.5) No Action (See NA NA (NTTF Recommendations) Technical Report APR1400- E-P-NR-7.2 Order licensees to provide safety-related ac electrical power 14005-P, Rev 2) for the spent fuel pool makeup system.

7.3 Order licensees to revise their technical specifications to No Action (See NA NA address requirements to have one train of on-site emergency Technical Report electrical power operable for spent fuel pool makeup and APR1400- E-P-NR-spent fuel pool instrumentation when there is irradiated fuel 14005-P, Rev 2) in the spent fuel pool, regardless of the operational mode of the reactor.

7.4 Order licensees to have an installed seismically qualified No Action (See NA NA means to spray water into the spent fuel pools, including an Technical Report easily accessible connection to supply the water (e.g., using a APR1400- E-P-NR-portable pump or pumper truck) at grade outside the building. 14005-P, Rev 2) 7.5 Initiate rulemaking or licensing activities or both to require No Action (See NA NA the actions related to the spent fuel pool described in detailed Technical Report recommendations 7.1-7.4. APR1400- E-P-NR-14005-P, Rev 2) 9.3 Emergency preparedness regulatory actions (the remaining portions No Action NA NA 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 1.9-106 Rev. 2

APR1400 DCD TIER 2 Table 1.9-8 (14 of 19)

NTTF Rec. NRC Recommendations/Requirements in SECY-11-0093, SECY- Applicable No 11-0137, SECY-12-0025, SECY-12-0095, EA-12-049, EA-12-051 APR1400 Design DCD Section COL Action Note 9.3 2. Develop and issue an order to address those changes (cont.) 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.

Tier 3 (Those NTTF Recommendations that require further staff study to support a regulatory action) 2.2 Ten-year confirmation of seismic and flooding hazards (dependent No Action - See NA NA on Recommendation 2.1) Technical Report Initiate rulemaking to require licensees to confirm seismic APR1400-E-P-NR-hazards and flooding hazards every 10 years and address any 14005-P, Rev 2 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 prevent or mitigate No Action - See NA NA seismically-induced fires and floods (long-term evaluation) Technical Report The Task Force recommends, as part of the longer term review, APR1400-E-P-NR-that the NRC evaluate potential enhancements to the capability to 14005-P, Rev 2 prevent or mitigate seismically induced fires and floods.

1.9-107 Rev. 2

APR1400 DCD TIER 2 Table 1.9-8 (15 of 19)

NTTF Rec. NRC Recommendations/Requirements in SECY-11-0093, SECY- Applicable No 11-0137, SECY-12-0025, SECY-12-0095, EA-12-049, EA-12-051 APR1400 Design DCD Section COL Action Note 5.2 Reliable hardened vents for other containment designs (long-term No Action - See NA NA evaluation) Technical Report Reevaluate the need for hardened vents for other containment APR1400-E-P-NR-designs, considering the insights from the Fukushima accident. 14005-P, Rev 2 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 containment or in other No Action - See NA NA buildings (long-term evaluation) Technical Report The Task Force recommends, as part of the longer term review, APR1400-E-P-NR-that the NRC identify insights about hydrogen control and 14005-P, Rev 2 mitigation inside containment or in other buildings as additional information is revealed through further study of the Fukushima Dai-ichi accident.

9.1 & Emergency preparedness (EP) enhancements for prolonged SBO and No Action - See NA NA 9.2 multiunit events (dependent on availability of critical skill sets) Technical Report 9.1 Initiate rulemaking to require EP enhancements for multiunit APR1400-E-P-NR-events in the following areas: 14005-P, Rev 2

• personnel and staffing

• dose assessment capability

• training and exercises

• equipment and facilities 1.9-108 Rev. 2

APR1400 DCD TIER 2 Table 1.9-8 (16 of 19)

NTTF Rec. NRC Recommendations/Requirements in SECY-11-0093, SECY- Applicable No 11-0137, SECY-12-0025, SECY-12-0095, EA-12-049, EA-12-051 APR1400 Design DCD Section COL Action Note 9.1 & 9.1 Initiate rulemaking to require EP enhancements for 9.2 prolonged SBO in the following areas:

(cont.)

• communications capability

• ERDS capability

• training and exercises

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

10) Technical Report Order licensees to do the following until rulemaking is complete: APR1400-E-P-NR-14005-P, Rev 2

• Maintain ERDS capability throughout the accident.

10 Additional EP topics for prolonged SBO and multiunit events (long- No Action - See NA NA term evaluation) Technical Report 10.1 Analyze current protective equipment requirements for APR1400-E-P-NR-emergency responders and guidance based upon insights 14005-P, Rev 2 from the accident at Fukushima.

1.9-109 Rev. 2

APR1400 DCD TIER 2 Table 1.9-8 (17 of 19)

NTTF Rec. NRC Recommendations/Requirements in SECY-11-0093, SECY- Applicable No 11-0137, SECY-12-0025, SECY-12-0095, EA-12-049, EA-12-051 APR1400 Design DCD Section COL Action Note 10 10.2 Evaluate the command and control structure and the (cont.) qualifications of decision-makers to ensure that the proper level of authority and oversight 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.

1.9-110 Rev. 2

APR1400 DCD TIER 2 Table 1.9-8 (18 of 19)

NTTF Rec. NRC Recommendations/Requirements in SECY-11-0093, SECY- Applicable No 11-0137, SECY-12-0025, SECY-12-0095, EA-12-049, EA-12-051 APR1400 Design DCD Section COL Action Note 11 EP topics for decision-making, radiation monitoring, and public No Action NA NA education (long-term evaluation) 11.1 Study whether enhanced on-site emergency response resources are necessary to support the effective implementation of the licensees 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 on-site 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.

1.9-111 Rev. 2

APR1400 DCD TIER 2 Table 1.9-8 (19 of 19)

NTTF Rec. NRC Recommendations/Requirements in SECY-11-0093, SECY- Applicable No 11-0137, SECY-12-0025, SECY-12-0095, EA-12-049, EA-12-051 APR1400 Design DCD Section COL Action Note 12.1 Reactor Oversight Process modifications to reflect the recommended No Action NA NA 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 Inspector Training No Action NA NA on SAMGs (dependent on Recommendation 8)

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

1.9-112 Rev. 2