ML18079A464

Korea Hydro-Nuclear Power Co., Ltd (Design Control Document) Rev.2 - Tier2 Chapter12-Radiation Protection

ML18079A464
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: ML18079A464 (380)

Text

APR1400 DESIGN CONTROL DOCUMENT TIER 2 CHAPTER 12 RADIATION PROTECTION 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 12 - RADIATION PROTECTION TABLE OF CONTENTS NUMBER TITLE PAGE CHAPTER 12 - RADIATION PROTECTION......................................................... 12.1-1 12.1 Ensuring that Occupational Radiation Exposures Are As Low As (Is) Reasonably Achievable .............................................................................. 12.1-1 12.1.1 Policy Considerations ........................................................................ 12.1-1 12.1.1.1 Design Policies ............................................................... 12.1-2 12.1.1.2 Operation Policies .......................................................... 12.1-3 12.1.2 Design Considerations ....................................................................... 12.1-5 12.1.2.1 General Design Considerations for Maintaining Occupational Exposures ALARA................................... 12.1-5 12.1.2.2 Equipment Design Considerations for Maintaining Occupational Exposures ALARA................................... 12.1-6 12.1.2.3 Facility Layout Design Considerations for Maintaining Occupational Exposures ALARA................................... 12.1-7 12.1.3 Operational Considerations ............................................................... 12.1-9 12.1.3.1 General Operational Considerations for Maintaining Exposures ALARA ......................................................... 12.1-9 12.1.3.2 Design Features for ALARA During Maintenance and Inspection ....................................................................... 12.1-9 12.1.3.3 Post-Accident Conditions ............................................. 12.1-11 12.1.3.4 Decommissioning ......................................................... 12.1-11 12.1.4 Combined License Information ....................................................... 12.1-12 12.1.5 References ........................................................................................ 12.1-12 12.2 Radiation Sources ............................................................................................. 12.2-1 12.2.1 Contained Sources ............................................................................. 12.2-1 12.2.1.1 Reactor Containment Building ....................................... 12.2-1 12.2.1.2 Auxiliary Building ........................................................ 12.2-13 12.2.1.3 Turbine Generator Building .......................................... 12.2-15 12.2.1.4 Compound Building ..................................................... 12.2-15 i Rev. 2

APR1400 DCD TIER 2 12.2.1.5 Sources Resulting from Design Basis Accidents .......... 12.2-17 12.2.1.6 Stored Radioactivity ..................................................... 12.2-17 12.2.1.7 Pipe Routing ................................................................. 12.2-18 12.2.2 Airborne Radioactive Material Sources ........................................... 12.2-18 12.2.2.1 Production of Airborne Radioactive Material .............. 12.2-19 12.2.2.2 Airborne Sources in Normally Accessible Areas ......... 12.2-19 12.2.2.3 Airborne Concentrations............................................... 12.2-20 12.2.2.4 Basis and Assumptions for Partition Factors ................ 12.2-21 12.2.3 Sources Used in NUREG-0737 Post-Accident Shielding Analysis ........................................................................................... 12.2-22 12.2.4 Combined License Information ....................................................... 12.2-22 12.2.5 References ........................................................................................ 12.2-23 12.3 Radiation Protection Design Features ............................................................ 12.3-1 12.3.1 Facility Design Features .................................................................... 12.3-1 12.3.1.1 General Arrangement Design Features ........................... 12.3-1 12.3.1.2 Equipment and System Design Features ........................ 12.3-5 12.3.1.3 Source Term Control..................................................... 12.3-14 12.3.1.4 Airborne Contamination Control .................................. 12.3-15 12.3.1.5 Equipment Improvements ............................................ 12.3-16 12.3.1.6 Radiation Zone Designation ......................................... 12.3-18 12.3.1.7 General Design Considerations to Keep Post-Accident Exposures ALARA ....................................................... 12.3-19 12.3.1.8 Post-Accident Radiation Zones .................................... 12.3-19 12.3.1.9 Vital Area Access .......................................................... 12.3-20 12.3.2 Shielding .......................................................................................... 12.3-23 12.3.2.1 General Shielding Design Criteria................................ 12.3-23 12.3.2.2 Shielding Analysis ........................................................ 12.3-25 12.3.2.3 Shielding Design .......................................................... 12.3-28 12.3.2.4 Access Control to High Radiation and Very High Radiation Areas ............................................................ 12.3-31 12.3.3 Ventilation ........................................................................................ 12.3-33 12.3.4 Area Radiation and Airborne Radioactivity Monitoring Instrumentation ................................................................................ 12.3-34 ii Rev. 2

APR1400 DCD TIER 2 12.3.4.1 Area Radiation Monitoring System .............................. 12.3-34 12.3.4.2 Airborne Radioactivity Monitoring Instrumentation.... 12.3-41 12.3.5 Dose Assessment ............................................................................. 12.3-42 12.3.6 Combined License Information ....................................................... 12.3-42 12.3.7 References ........................................................................................ 12.3-44 12.4 Dose Assessment and Minimization of Contamination ................................. 12.4-1 12.4.1 Dose Assessment ............................................................................... 12.4-1 12.4.1.1 Estimated Annual Occupational Exposures.................... 12.4-1 12.4.1.2 Post-Accident Actions .................................................... 12.4-9 12.4.1.3 Estimated Annual Dose at the Exclusion Area Boundary ...................................................................... 12.4-10 12.4.1.4 Estimated Dose to Construction Workers ..................... 12.4-11 12.4.2 Minimization of Contamination and Radioactive Waste Generation........................................................................................ 12.4-11 12.4.2.1 Design Considerations .................................................. 12.4-12 12.4.2.2 Operational/Programmatic Considerations................... 12.4-15 12.4.2.3 Summary of Design Features ....................................... 12.4-16 12.4.2.4 Design Details to Address Minimization of Contamination .............................................................. 12.4-16 12.4.3 Combined License Information ....................................................... 12.4-18 12.4.4 References ........................................................................................ 12.4-19 12.5 Operational Radiation Protection Program................................................... 12.5-1 12.5.1 Combined License Information ......................................................... 12.5-2 12.5.2 References .......................................................................................... 12.5-2 iii Rev. 2

APR1400 DCD TIER 2 LIST OF TABLES NUMBER TITLE PAGE Table 12.2-1 Normal Operation Neutron Spectra Outside the Reactor Vessel ....... 12.2-24 Table 12.2-2 Normal Operation Gamma Spectra Outside the Reactor Vessel ....... 12.2-25 Table 12.2-3 Shutdown Gamma Spectra Outside the Reactor Vessel .................... 12.2-26 Table 12.2-4 Basis for Reactor Coolant Fission Product Activities ........................ 12.2-27 Table 12.2-5 Reactor Coolant Equilibrium Concentration (Core Power: 4,063 MWt, 0.25 % Fuel Defect, No Gas Stripping) ................................... 12.2-28 Table 12.2-6 Deposited Crud Activities (SG Tubing and Reactor System Piping) ................................................................................................ 12.2-30 Table 12.2-7 N-16 Activity ..................................................................................... 12.2-31 Table 12.2-8 Parameters Used in Spent Fuel Decay Gamma Source Calculation ......................................................................................... 12.2-32 Table 12.2-9 Spent Fuel Gamma Source ................................................................. 12.2-33 Table 12.2-10 CVCS Heat Exchanger Inventories, Maximum Values (Bq)............. 12.2-34 Table 12.2-11 CVCS Ion Exchanger Inventories, Maximum Values (Bq) ............... 12.2-36 Table 12.2-12 CVCS Filter Inventories, Maximum Values (Bq).............................. 12.2-38 Table 12.2-13 CVCS Tank Inventories, Maximum Values (Bq) .............................. 12.2-40 Table 12.2-14 Boric Acid Concentrator Package Inventories, Maximum Values (Bq) .................................................................................................... 12.2-45 Table 12.2-15 Gamma Sources of Irradiated ICI, CEA, NSA, and SCA.................. 12.2-47 Table 12.2-16 Shutdown Cooling System Specific Source Strength ........................ 12.2-48 Table 12.2-17 Fission and Corrosion Product Activities in the Spent Fuel Pool ...... 12.2-49 Table 12.2-17a SFP Demineralizer and Filter Source Terms ..................................... 12.2-50 Table 12.2-18 Steam Generator Blowdown and Condensate Polishing System Source Terms (0.25 % Fuel Defect) ................................................... 12.2-51 Table 12.2-19 Gaseous Radwaste System Source Terms (0.25 % Fuel Defect) ....... 12.2-54 Table 12.2-20 Liquid Radwaste System Tank Source Terms (Bq/cm3).................... 12.2-56 Table 12.2-21 Liquid Radwaste System Component Source Terms (Bq) ................ 12.2-59 Table 12.2-22 Solid Radwaste System Tank Source Terms (Bq) ............................. 12.2-62 iv Rev. 2

APR1400 DCD TIER 2 Table 12.2-22a Waste Drum Storage Area Source Terms (Bq) .................................. 12.2-63 Table 12.2-23 Airborne Radioactivity Concentrations.............................................. 12.2-64 Table 12.2-24 Source Terms for Post-Accident Shielding Analysis ......................... 12.2-68 Table 12.2-25 Radioactive Source Dimensions and Parameters Used in Shielding Analysis ............................................................................. 12.2-82 Table 12.2-26 Assumptions and Parameters Used in Airborne Source Term Calculations ........................................................................................ 12.2-87 Table 12.2-27 Vapor Activities in Pressurizer .......................................................... 12.2-95 Table 12.2-28 Parameters Used in CVCS Inventory Calculations ............................ 12.2-96 Table 12.3-1 Cobalt Contents of Finished Surface for Primary System Materials ............................................................................................. 12.3-47 Table 12.3-1a Cobalt Contents of Fuel Assembly Components Materials ............... 12.3-48 Table 12.3-2 Normal Operation Radiation Zone Designations ............................... 12.3-49 Table 12.3-3 Post-Accident Radiation Zone Designations and Design Dose Rates ................................................................................................... 12.3-50 Table 12.3-4 Design Basis Radiation Shield Thicknesses ...................................... 12.3-51 Table 12.3-5 Areas Potentially Greater than 1 Gy/hr .............................................. 12.3-67 Table 12.3-6 Area Radiation Monitors .................................................................... 12.3-69 Table 12.4-1 Occupational Dose Estimates During Reactor Operations and Surveillance ........................................................................................ 12.4-21 Table 12.4-2 Occupational Dose Estimates During Routine Maintenance ............. 12.4-22 Table 12.4-3 Occupational Dose Estimates During Inservice Inspection ............... 12.4-23 Table 12.4-4 Occupational Dose Estimates During Special Maintenance .............. 12.4-24 Table 12.4-5 Occupational Dose Estimates During Waste Processing ................... 12.4-25 Table 12.4-6 Occupational Dose Estimates During Refueling................................ 12.4-26 Table 12.4-7 Annual Personnel Doses per Activity Categories .............................. 12.4-27 Table 12.4-8 Estimated Accident Mission Dose ..................................................... 12.4-28 Table 12.4-9 APR1400 Plant Systems Requiring Full NRC RG 4.21 Evaluations ......................................................................................... 12.4-49 v Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 NRC RG 4.21 Design Objectives and Applicable DCD Subsection Information for Minimizing Contamination and Generation of Radioactive Waste....................................................... 12.4-51 vi Rev. 2

APR1400 DCD TIER 2 LIST OF FIGURES NUMBER TITLE PAGE Figure 12.2-1 Spent Fuel Assembly Dose Rate vs. Axial Distance in Refueling Pool .................................................................................................... 12.2-98 Figure 12.2-2 Spent Fuel Assembly Dose Rate vs. Radial Distance in Refueling Pool .................................................................................... 12.2-99 Figure 12.3-1 Radiation Zones(Normal) Auxiliary/Containment Building El.55'-0".............................................................................................. 12.3-71 Figure 12.3-2 Radiation Zones (Normal) Auxiliary Building - Partial Plan El. 68'-0", 77'-0", and 86'-0" .............................................................. 12.3-72 Figure 12.3-3 Radiation Zones (Normal) Auxiliary/Containment Building El.78'-0".............................................................................................. 12.3-73 Figure 12.3-4 Radiation Zones(Normal) Auxiliary/Containment Building El.100'-0"............................................................................................ 12.3-74 Figure 12.3-5 Radiation Zones(Normal) Auxiliary/Containment Building El.120'-0"............................................................................................ 12.3-75 Figure 12.3-6 Radiation Zones(Normal) Auxiliary/Containment Building El.137'-6"............................................................................................ 12.3-76 Figure 12.3-7 Radiation Zones(Normal) Auxiliary/Containment Building El.156'-0"............................................................................................ 12.3-77 Figure 12.3-8 Radiation Zones(Normal) Auxiliary/Containment Building El.174'-0"............................................................................................ 12.3-78 Figure 12.3-9 Radiation Zones (Normal) Auxiliary/Containment Building El.

195'-0" and Roof Plan ........................................................................ 12.3-79 Figure 12.3-10 Radiation Zones (Normal) Compound Building El. 63'-0"................ 12.3-80 Figure 12.3-11 Radiation Zones (Normal) Compound Building Partial Plan El. 77-0 ............................................................................................ 12.3-81 Figure 12.3-12 Radiation Zones (Normal) Compound Building El. 85'-0"................ 12.3-82 Figure 12.3-13 Radiation Zones (Normal) Compound Building El. 100'-0".............. 12.3-83 Figure 12.3-14 Radiation Zones (Normal) Compound Building El. 120'-0".............. 12.3-84 Figure 12.3-15 Radiation Zones (Normal) Compound Building El. 139'-6".............. 12.3-85 vii Rev. 2

APR1400 DCD TIER 2 Figure 12.3-16 Radiation Zones (Normal) Compound Building El. 157'-9" &

Roof Plan ............................................................................................ 12.3-86 Figure 12.3-17 Radiation Zones (Normal) Turbine Building El. 73'-0" ..................... 12.3-87 Figure 12.3-18 Radiation Zones (Normal ) Turbine Building El. 100'-0" .................. 12.3-88 Figure 12.3-19 Radiation Zones (Normal) CCW HX Building .................................. 12.3-89 Figure 12.3-20 Radiation Zones (1 Hour after Accident)-Auxiliary/Containment Building El. 55'-0".............................................................................. 12.3-90 Figure 12.3-21 Radiation Zones (1 Day after Accident)-Auxiliary/Containment Building El. 55'-0".............................................................................. 12.3-91 Figure 12.3-22 Radiation Zones (1 Week after Accident)-

Auxiliary/Containment Building El. 55'-0" ....................................... 12.3-92 Figure 12.3-23 Radiation Zones (1 Month after Accident)-

Auxiliary/Containment Building El. 55'-0" ....................................... 12.3-93 Figure 12.3-24 Radiation Zones (1 Hour after Accident)-Auxiliary/Containment Building El. 78'-0".............................................................................. 12.3-94 Figure 12.3-25 Radiation Zones (1 Day after Accident)-Auxiliary/Containment Building El. 78'-0".............................................................................. 12.3-95 Figure 12.3-26 Radiation Zones (1 Week after Accident)-

Auxiliary/Containment Building El. 78'-0" ....................................... 12.3-96 Figure 12.3-27 Radiation Zones (1 Month after Accident)-

Auxiliary/Containment Building El. 78'-0" ....................................... 12.3-97 Figure 12.3-28 Radiation Zones (1 Hour after Accident)-Auxiliary/Containment Building El. 100'-0"............................................................................ 12.3-98 Figure 12.3-29 Radiation Zones (1 Day after Accident)-Auxiliary/Containment Building El. 100'-0"............................................................................ 12.3-99 Figure 12.3-30 Radiation Zones (1 Week after Accident)-

Auxiliary/Containment Building El. 100'-0" ................................... 12.3-100 Figure 12.3-31 Radiation Zones (1 Month after Accident)-

Auxiliary/Containment Building El. 100'-0" ................................... 12.3-101 Figure 12.3-32 Radiation Zones (1 Hour after Accident)-Auxiliary/Containment Building El. 120'-0".......................................................................... 12.3-102 Figure 12.3-33 Radiation Zones (1 Day after Accident)-Auxiliary/Containment Building El. 120'-0".......................................................................... 12.3-103 viii Rev. 2

APR1400 DCD TIER 2 Figure 12.3-34 Radiation Zones (1 Week after Accident)-

Auxiliary/Containment Building El. 120'-0" ................................... 12.3-104 Figure 12.3-35 Radiation Zones (1 Month after Accident)-

Auxiliary/Containment Building El. 120'-0" ................................... 12.3-105 Figure 12.3-36 Radiation Zones (1 Hour after Accident)-Auxiliary/Containment Building El. 137'-6".......................................................................... 12.3-106 Figure 12.3-37 Radiation Zones (1 Day after Accident)-Auxiliary/Containment Building El. 137'-6".......................................................................... 12.3-107 Figure 12.3-38 Radiation Zones (1 Week after Accident)-

Auxiliary/Containment Building El. 137'-6" ................................... 12.3-108 Figure 12.3-39 Radiation Zones (1 Month after Accident)-

Auxiliary/Containment Building El. 137'-6" ................................... 12.3-109 Figure 12.3-40 Radiation Zones (1 Hour after Accident)-Auxiliary/Containment Building El. 156'-0".......................................................................... 12.3-110 Figure 12.3-41 Radiation Zones (1 Day after Accident)-Auxiliary/Containment Building El. 156'-0".......................................................................... 12.3-111 Figure 12.3-42 Radiation Zones (1 Week after Accident)-

Auxiliary/Containment Building El. 156'-0" ................................... 12.3-112 Figure 12.3-43 Radiation Zones (1 Month after Accident)-

Auxiliary/Containment Building El. 156'-0" ................................... 12.3-113 Figure 12.3-44 Radiation Zones (1 Hour after Accident)-Auxiliary/Containment Building El. 174'-0".......................................................................... 12.3-114 Figure 12.3-45 Radiation Zones (1 Day after Accident)-Auxiliary/Containment Building El. 174'-0".......................................................................... 12.3-115 Figure 12.3-46 Radiation Zones (1 Week after Accident)-

Auxiliary/Containment Building El. 174'-0" ................................... 12.3-116 Figure 12.3-47 Radiation Zones (1 Month after Accident)-

Auxiliary/Containment Building El. 174'-0" ................................... 12.3-117 Figure 12.3-48 Radiation Zones (1 Hour after Accident)-Auxiliary/Containment Building El. 195'-0" & Roof Plan .................................................... 12.3-118 Figure 12.3-49 Radiation Zones (1 Day after Accident)-Auxiliary/Containment Building El. 195'-0" & Roof Plan .................................................... 12.3-119 Figure 12.3-50 Radiation Zones (1 Week after Accident)-

Auxiliary/Containment Building El. 195'-0" & Roof Plan .............. 12.3-120 ix Rev. 2

APR1400 DCD TIER 2 Figure 12.3-51 Radiation Zones (1 Month after Accident)-

Auxiliary/Containment Building El. 195'-0" & Roof Plan .............. 12.3-121 Figure 12.3-52 Sectional View of Fuel Transfer Tube ............................................. 12.3-122 Figure 12.4-1 Typical Design Details of Drainage and Leak Detection Configuration ................................................................................... 12.4-126 Figure 12.4-2 Typical Design Details of Leak Detection Instrumentation ............. 12.4-127 Figure 12.4-3 Typical Compound Building Trench Location Schematic............... 12.4-128 Figure 12.4-4 Typical Design for Building Sleeves Schematic .............................. 12.4-129 x Rev. 2

APR1400 DCD TIER 2 ACRONYM AND ABBREVIATION LIST AB auxiliary building AC alternation current AFW auxiliary feedwater APR1400 Advanced Power Reactor 1400 AST alternative source term BOC beginning of cycle CCW component cooling water CEACP control element assembly change platform CEA control element assembly CEDE committed effective dose equivalent CLVPS containment low-volume purge system COL combined license CPIAS containment purge isolation actuation signal CS containment spray CSS containment spray system CVCS chemical and volume control system DBA design basis accident DCF dose conversion factor DDE deep dose equivalent DE dose equivalent DF decontamination factor EAB exclusion area boundary EDE effective dose equivalent EFPD effective full-power day EOC end of cycle ESF engineered safety features ESFAS engineered safety features actuation system FF flash fraction GDC general design criteria xi Rev. 2

APR1400 DCD TIER 2 HEPA high-efficiency particulate air HI hydrogen igniter HVAC heating, ventilating, and air conditioning ICRP International Commission on Radiological Protection IRWST in-containment refueling water storage tank LOCA loss-of-coolant accident MCR main control room MSIV main steam isolation valve NRC United States Nuclear Regulatory Commission PA postulated accident POSRV pilot operated safety and relief valve PTS primary-to-secondary PWR pressurized water reactor RADTRAD radionuclide transport, removal, and dose RCFC reactor containment fan cooler RCGVS reactor coolant gas vent system RCP reactor coolant pump RCPB reactor coolant pressure boundary RCS reactor coolant system RMS radiation monitoring system RV reactor vessel SCS shutdown cooling system SFP spent fuel pool SG steam generator SGBS steam generator blowdown system SI safety injection SIS safety injection system TEDE total effective dose equivalent TSC technical support center xii Rev. 2

APR1400 DCD TIER 2 CHAPTER 12 - RADIATION PROTECTION 12.1 Ensuring that Occupational Radiation Exposures Are As Low As (Is)

Reasonably Achievable This chapter describes the radiation protection measures and operating policies for the APR1400 that meet the radiation protection standards in 10 CFR Part 20 and U.S. Nuclear Regulatory Commission (NRC) guidance related to protection from radiation, including the internal and external radiation that plant operating and construction personnel and the public are potentially exposed to during normal operation, including anticipated operational occurrences (AOOs). The combined license (COL) applicant is to supplement the radiation protection measures and operating policies by developing site-specific procedures and programs related to radiation protection.

Radiation protection and health physics are used interchangeably in this chapter.

The radiation protection measures in the design include the separation of radioactive components into individual shielded compartments, shielding that is designed to adequately attenuate radiation emitted from the components and piping that are sources of significant ionizing radiation, remote operation of equipment, ventilation of equipment areas with the potential to have airborne radiation, permanent radiation monitoring systems (RMSs),

training of personnel in radiation protection, and administrative policies and procedures to maintain exposures ALARA.

The radiation protection program for the APR1400 associated with maintaining occupational exposure to radiation as low as reasonably achievable (ALARA) is described below, and conforms with Nuclear Energy Institute (NEI)07-08A (Reference 1).

12.1.1 Policy Considerations The design policy of the APR1400 is to maintain occupational radiation exposure ALARA to operating and construction personnel and to the public. 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 Regulatory Guides (RGs) 1.8 (Reference 2), 1.33 (Reference 3), 8.8 (Reference 4), and 8.10 (Reference 5) (COL 12.1(1)). The ALARA policy is applied to the total exposures (person-Sv) received by all operation and construction personnel, and to individual 12.1-1 Rev. 2

APR1400 DCD TIER 2 exposures. The proper approach and awareness of potential problems related to radiation protection are addressed through the proper training of all plant personnel as provided in NRC RG 1.8.

12.1.1.1 Design Policies The design policy of the APR1400 is to implement the ALARA philosophy during the early stages of the design by reviewing and documenting the ALARA design to provide reasonable assurance of consistency with the applicable ALARA design criteria. The design policy is based on experience, including lessons learned, which indicates that the most effective design approach is to implement ALARA criteria during the early design stages to minimize the need for design changes in later stages.

During the early development of the design, the system engineers and designers are trained in the ALARA design principles and in radiation protection measures through procedure review assignments and periodic group training. ALARA design reviews are periodically conducted throughout the design and construction phases to update and modify the design based on the ongoing development of the plants detailed design and lessons learned from operating plants. The reviews are conducted to integrate the layout, shielding, ventilation, and radiation monitoring design with security, access control, maintenance, and inservice inspection (ISI) to provide reasonable assurance that occupational exposures are ALARA.

Onsite inspections are conducted during construction to provide reasonable assurance that the shielding, components, and piping layout are built in accordance with the established design. During construction, ALARA-experienced personnel conduct visual inspections to provide reasonable assurance that there are no instances of uncontrolled radiation, such as scattering and unforeseen streaming paths due to the locations of piping penetrations in the shield walls as they are installed.

Subsections 12.1.2 and 12.1.3 describe the criteria and methods used in the design and operation of shielding, ventilation, and radiation monitoring systems (including equipment and plant arrangements and access control provisions) to keep occupational exposures ALARA.

The design of the radiation protection measures is reviewed by qualified engineers as a normal part of the plant design process. The personnel responsible for ALARA design and review are engineers who are competent in the application of radiation protection 12.1-2 Rev. 2

APR1400 DCD TIER 2 principles, including radiological dose assessment, shielding design, and radwaste system design.

The radiation protection team reviews operating plant exposures and designs to establish criteria and guidelines for radiation protection. The team reviews various aspects of the plant design to ascertain conformance with the established criteria of maintaining exposures ALARA and recommends design modifications as necessary. During construction, field changes submitted to the plant designer are reviewed by radiation protection engineers to provide reasonable assurance that occupational exposures are ALARA. The reviews meet the intent of NRC RG 8.8.

12.1.1.2 Operation Policies The APR1400 design is supplemented by operation policies and programs that are intended to keep occupational exposure to radiation ALARA while satisfying the applicable regulations. The program consists of written procedures for executing maintenance, testing, and inspection activities in radiation areas and training for plant personnel in radiation protection fundamentals that is conducted periodically and reviewed by plant management.

The general manager of the radiation protection team, who is responsible for the initial establishment of the health physics program, reports to the plant manager and exercises supervisory control over the radiation protection team. Radiation protection technicians are responsible for radiation surveys, contamination surveys, and environmental sampling.

The functions of the radiation protection team include the following:

a. Prepare and approve detailed procedures for radiation protection before plant operation begins and provide reasonable assurance that dose limits are established in accordance with the approved health physics program and all applicable regulations

b. Provide reasonable assurance that radiation exposure to plant personnel is maintained ALARA by:

1) Evaluating radiological conditions in areas to be accessed by personnel and taking precautionary measures to provide reasonable assurance of minimal exposure 12.1-3 Rev. 2

APR1400 DCD TIER 2

2) Controlling personnel and equipment movement into and out of radiation control areas

3) Providing instruction and supervision for the proper use and care of special protective clothing and equipment

4) Posting each radiation control area with conspicuous and appropriate caution signs, and access control means (locking devices, rope) in accordance with the applicable regulations

5) Administering and controlling conditions of radiation work permits for work in areas with high radiation or contamination levels in accordance with the approved procedures and the radiation protection program

c. Determine procedural requirements for the use of personnel monitoring devices and maintain records of personnel exposure in accordance with the applicable regulations

d. Control and document all radioactive material entering or leaving the plant site in accordance with the applicable regulations

e. Establish procedures for dealing with emergency conditions

f. Train plant staff and visitors in the radiation protection policies and procedures, as required

g. Include a dosimetry program as part of the site health physics program. The program includes whole-body counting to measure the uptake of radiation by personnel

h. Conduct periodic calibration for process radiation, area radiation, portable radiation, and airborne radioactivity monitors

i. Monitor and maintain records of all radioactive effluents released from the station

j. Collect data required in the evaluation of radiological effects from plant operation 12.1-4 Rev. 2

APR1400 DCD TIER 2 12.1.2 Design Considerations This section describes the general design features included in the plant and system design to implement the policies presented in Subsection 12.1.1. Provisions and design features for maintaining personnel exposures ALARA are presented in Section 12.3.

12.1.2.1 General Design Considerations for Maintaining Occupational Exposures ALARA Design guidance to maintain occupational exposures ALARA is provided in the APR1400 ALARA Design Guide, which describes the design approach and operational implementation to reduce radiation exposure to plant personnel, including maintenance personnel. The APR1400 ALARA Design Guide is consistent with the recommendations in NRC RG 8.8 and NRC RG 8.10. It also incorporates lessons learned from earlier nuclear plant designs and operating experience that are obtained by reviewing and evaluating applicable operating experience, updated industry practices, and implementation of improvement through internal lessons-learned programs.

The APR1400 ALARA Design Guide also provides the procedure for training the responsible design engineers in each design discipline including the process system; plant layout; piping; heating, ventilating, and air conditioning (HVAC); instrumentation and control (I&C); and civil/structure during the early stage of the plant design through periodic procedure review and documented group training sessions. The independent ALARA review process is performed by experienced radiation protection engineers to provide reasonable assurance that the ALARA principles are implemented.

Experience from existing nuclear power plant designs and operating pressurized water reactor (PWR) plants in Korea and the United States is incorporated into the APR1400 design process to provide reasonable assurance that occupational radiation exposure (ORE) is maintained ALARA. Because Korea Hydro & Nuclear Power Co., Ltd. (KHNP) has constructed or is constructing and operating all of the Korean nuclear power plants, the feedback and lessons learned from past construction and operation are documented and incorporated in the design of the new plant. A lessons-learned program is used so that NRC generic communications and operating and construction experience are reviewed, assessed, and implemented as appropriate.

The ALARA design criteria in the APR1400 ALARA Design Guide are established to accomplish the following design objectives:

12.1-5 Rev. 2

APR1400 DCD TIER 2

a. To reduce the duration and/or frequency of required equipment maintenance through the use of industry-proven equipment and construction materials that are commensurate with the intended plant processes and services

b. To reduce radiation levels through the use of operational controls such as frequent flushing and cleaning of equipment and the use of localized shielding where needed

c. To reduce the time spent in radiation areas where maintenance and other operational activities are necessary by providing easy access to components that require frequent maintenance and by moving components to low-radiation zones to perform maintenance when possible The APR1400 design features that meet the above design criteria are described in Subsections 12.3.1.1 and 12.3.1.2.

The APR1400 is also designed to minimize the production, distribution, and retention of activated corrosion products throughout the reactor coolant system (RCS). The production of corrosion products is minimized by the proper selection of materials and the implementation of chemistry control of the primary system, as described in Subsection 12.3.1.3. Reactor coolant leakage is controlled to minimize the distribution and retention of corrosion products as described in Subsection 12.3.1.2(i).

In accordance with the requirements in 10 CFR 20.1406 (Reference 6), the APR1400 design implements the design approaches to minimize, to the extent practicable, contamination of the facility and the environment; facilitates eventual decommissioning; and minimizes, to the extent practicable, the generation of radioactive waste. The design considerations that conform with these requirements are described in Subsection 12.4.2.

12.1.2.2 Equipment Design Considerations for Maintaining Occupational Exposures ALARA NRC RG 8.8 provides guidance for the selection of equipment that has an integral part in maintaining occupational exposures ALARA. In accordance with NRC RG 8.8, the selection of APR1400 equipment is based on (1) the effectiveness of contamination removal through the use of improved technologies, (2) a reduced need for equipment maintenance through the use of industry-proven equipment with demonstrated reliability and the use of preventive maintenance chemistry control, and (3) material properties that 12.1-6 Rev. 2

APR1400 DCD TIER 2 are suitable for minimizing corrosion. The use of materials containing cobalt is minimized to the greatest extent possible.

Components with the potential for crud accumulation are designed to prevent settlement of solids through the use of ellipsoidal bottoms with internal spargers and flushing using either demineralized water or the chemical decontamination system. The criteria used for equipment selection are provided in the APR1400 ALARA Design Guide.

The APR1400 ALARA design considerations with respect to the equipment design are provided in Subsection 12.3.1.2 and summarized as follows:

a. Enhanced reliability of equipment that reduces the frequency of maintenance and thus the occupational exposures to workers during maintenance The use of reliable lamps with extended service time in high-radiation areas reduces the need for frequent lamp replacement.

Ion exchange and reverse osmosis are used to remove radioactive contamination.

Both methods have resulted in higher contaminant removal and more reliable operations than processes using radwaste evaporators.

b. Enhanced environmental qualification of equipment Equipment qualification for a variety of environmental conditions, such as radiation, humidity, and temperature, is addressed in Section 3.11. Electrical components containing radiation-sensitive materials are shielded or located in low-radiation areas.

c. Material selection of piping and components Material with low cobalt and antimony impurities is selected to minimize the production of corrosion products that are a significant contributor to occupational exposures during maintenance and operational activities.

12.1.2.3 Facility Layout Design Considerations for Maintaining Occupational Exposures ALARA The facility layout is designed so that the duration of time spent by workers within radiation areas for the performance of testing and maintenance activities is minimized.

12.1-7 Rev. 2

APR1400 DCD TIER 2 Also, the separation of more highly radioactive equipment from less radioactive equipment is implemented to the greatest extent practicable in the layout design.

The plant layout is designed to maintain occupational exposures ALARA during normal operation including AOOs, post-accident conditions, and decommissioning as follows:

a. Adequate spacing to facilitate the accessibility of equipment during maintenance activities and to enable the installation of temporary shielding, where necessary, during the performance of testing and maintenance activities in radiation areas.

b. Separation of nonradioactive systems from radioactive systems to the extent practicable to minimize the spread of contamination.

c. Installation of system components into separate cubicles or compartments based on their respective frequency of maintenance, operational characteristics, level of radioactivity, and inclusion of passive versus active components.

d. Design of the ventilation systems so that airflows from clean areas to potentially contaminated areas to minimize the potential for the spread of airborne contamination. For areas with the potential for high airborne radioactivity, a slightly negative pressure is maintained.

Equipment and piping containing radioactive fluids are arranged in accordance with their contamination levels and the ALARA design criteria to provide reasonable assurance that equipment and piping expected to contain radiation sources are adequately shielded and properly routed to minimize occupational exposures. The APR1400 ALARA design features with respect to plant layout are described in Subsection 12.3.1.1 and are summarized as follows:

a. Tanks are located in individual cubicles to the extent practicable and based on their levels of contamination; shield walls are sized adequately and conservatively to attenuate radiation and maintain low dose rates in areas occupied by plant personnel.

b. Pumps and valve galleries are provided outside the cubicles containing the radioactive equipment to minimize operational exposures during operation and maintenance activities.

12.1-8 Rev. 2

APR1400 DCD TIER 2

c. Instruments are located in low-radiation areas, and remote operation is implemented to the maximum extent practicable.

d. Piping that contains contaminated fluids is routed inside the pipe chases to the maximum extent practicable.

e. Shield wall penetrations are positioned to minimize radiation streaming from radioactive equipment to low-radiation areas.

12.1.3 Operational Considerations 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(2)). The APR1400 radiation protection design is implemented in conjunction with operational procedures and programs. Radiological health and safety procedures are developed and are continually reviewed and updated as necessary to provide reasonable assurance that occupational radiation exposures are ALARA. These procedures incorporate experience gained from plant operations. Improvements suggested during design and operation are incorporated and implemented to continually update the program to maximize protection of the plant personnel, the public, the environment, and the equipment.

The general manager of the radiation protection team is responsible for developing and updating radiological health and safety procedures. The plant manager supervises activities governed by these procedures so that workers are properly protected at all times.

All personnel receive the training required to perform their assigned tasks safely.

12.1.3.1 General Operational Considerations for Maintaining Exposures ALARA The COL applicant is to describe how the plant follows the guidance provided in NRC RGs 8.2 (Reference 7), 8.4 (Reference 25), 8.7 (Reference 8), 8.9 (Reference 9), 8.13 (Reference 10), 8.15 (Reference 11), 8.20 (Reference 12), 8.25 (Reference 13), 8.26 (Reference 14),

8.27 (Reference 15), 8.28 (Reference 16), 8.29 (Reference 17), 8.34 (Reference 18), 8.35 (Reference 19), 8.36 (Reference 20), and 8.38 (Reference 21) (COL 12.1(3)).

12.1.3.2 Design Features for ALARA During Maintenance and Inspection 12.1-9 Rev. 2

APR1400 DCD TIER 2 The APR1400 design incorporates lessons learned from past designs, as well as the recommendations from NRC RGs 8.8 and 8.10. Design features to provide reasonable assurance that personnel exposure is maintained ALARA during maintenance and inspection activities are as follows:

a. Platforms are provided around the steam generator (SG) to facilitate the handling of the reactor coolant pump (RCP) seal cartridges.

b. Sufficient spacing is provided for ingress and egress, equipment laydown/pull areas, and removal paths for major equipment that is expected to be replaced during the plant lifetime.

c. Removable insulation is used.

d. The length of welds is reduced through the use of seamless piping as much as possible.

e. An integrated head assembly (IHA) is used.

High-radiation areas that house systems and components requiring ISI are designed to facilitate worker ingress and egress. The APR1400 is also designed with adequate spacing, including laydown areas, for ingress and egress between equipment (to facilitate maintenance), and spacing for inspections. Radiation protection technicians survey areas that will be accessed for maintenance and inspection prior to worker entry and conduct periodic inspections during work activities.

The APR1400 also provides the following design features to maintain personnel exposure ALARA during inservice inspection:

a. Pipe stops, snubbers, and pipe hangers near welds that require ISI or periodic maintenance are carefully positioned to facilitate weld accessibility.

b. Integrally forged components and seamless piping are selected whenever possible to minimize the need for inservice weld inspections.

c. Blanket-type thermal insulation with hook-and-loop fasteners is selected for components and piping containing radioactive material. This type of insulation can be removed easily by one worker for inservice weld inspection.

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APR1400 DCD TIER 2

d. Adequate containment or ventilation is provided to minimize the potential for the spread of airborne contamination. Subsection 12.2.2 specifies the methodology for determining the in-plant concentration of airborne contamination. In Subsection 12.3.1.4, more information on airborne contamination control is described.

Following the radiation protection program, based on the guidance of NEI 07-03A (Reference 23), maintenance activities that could involve significant radiation exposure to personnel are carefully planned following the plant procedures to provide reasonable assurance of worker safety. Radiation work permits (RWPs) for routine and non-routine operations are issued for each activity to be performed in a high-radiation area. RWPs contain a list of the radiation protection requirements that are followed by all personnel working in a radiation area. Where practicable, radiation exposure reduction techniques, such as those in NRC RG 8.8, are evaluated and used.

12.1.3.3 Post-Accident Conditions The layout and system design includes considerations for post-accident conditions.

Separation and partitioning of components in individual shielded cubicles, provisions of adequate spacing, and ingress and egress design facilitate personnel access for surveillance and inspection, sample collection, cleanup, and repairs with minimum radiation exposure.

The design also facilitates the use of temporary shielding, such as the use of lead blankets, to further minimize radiation exposure.

12.1.3.4 Decommissioning The ALARA considerations for decommissioning are also incorporated into the system design and component layout through the provisions of individual shielded cubicles, adequate spacing, and ingress and egress design, and the partitioning and separation of low-contamination versus high-contamination areas. Other considerations for decommissioning include the provisions of decontamination capability to reduce residual contaminations, components designed with adequate capacities to reduce processing time, and the approach to maximize the use of integrated component assemblies to reduce stay time for dismantlement. In addition, decommissioning considerations for conformance with NRC RG 4.21 (Reference 22) and for applying to NEI 08-08A (Reference 24) are built into the design for systems and components that handle contaminated or potentially contaminated fluids. Decommissioning considerations are provided in Subsection 12.4.2.

12.1-11 Rev. 2

APR1400 DCD TIER 2 12.1.4 Combined License Information 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 1.8, 1.33, 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.

12.1.5 References

1. NEI 07-08A, Generic FSAR Template Guidance for Ensuring that Occupational Radiation Exposures are as Low as is Reasonably Achievable (ALARA), Nuclear Energy Institute, October 2009.

2. Regulatory Guide 1.8, Qualification & Training of Personnel for Nuclear Power Plants, Rev. 3, U.S. Nuclear Regulatory Commission, May 2000.

3. Regulatory Guide 1.33, Quality Assurance Program Requirements Operation, Rev. 3, U.S. Nuclear Regulatory Commission, June 2013.

4. Regulatory Guide 8.8, Information Relevant to Ensuring the Occupational Radiation Exposures at Nuclear Power Stations will be ALARA, Rev. 3, U.S. Nuclear Regulatory Commission, June 1978.

5. Regulatory Guide 8.10, Operating Philosophy for Maintaining Occupational Radiation Exposures As Low As Is Reasonably Achievable, Rev. 1-R, U.S. Nuclear Regulatory Commission, May 1977.

6. 10 CFR 20.1406, Minimization of Contamination, U.S. Nuclear Regulatory Commission.

12.1-12 Rev. 2

APR1400 DCD TIER 2

7. Regulatory Guide 8.2, Guide for Administrative Practices in Radiation Monitoring, Rev. 1, U.S. Nuclear Regulatory Commission, May 2011.

8. Regulatory Guide 8.7, Instructions for Recording and Reporting Occupational Radiation Exposure Data, Rev. 2, U.S. Nuclear Regulatory Commission, November 2005.

9. Regulatory Guide 8.9, Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program, Rev. 1, U.S. Nuclear Regulatory Commission, July 1993.

10. Regulatory Guide 8.13, Instruction Concerning Prenatal Radiation Exposure, Rev. 3, U.S. Nuclear Regulatory Commission, June 1999.

11. Regulatory Guide 8.15, Acceptable Programs for Respiratory Protection, Rev. 1, U.S. Nuclear Regulatory Commission, October 1999.

12. Regulatory Guide 8.20, Applications of Bioassay for I-125 and I-131, Rev. 1, U.S.

Nuclear Regulatory Commission, September 1979.

13. Regulatory Guide 8.25, Air Sampling in the Workplace, Rev. 1, U.S. Nuclear Regulatory Commission, June 1992.

14. Regulatory Guide 8.26, Applications of Bioassay for Fission and Activation Products, U.S. Nuclear Regulatory Commission, September 1980.

15. Regulatory Guide 8.27, Radiation Protection Training for Personnel at Light-Water-Cooled Nuclear Power Plants, U.S. Nuclear Regulatory Commission, March 1981.

16. Regulatory Guide 8.28, Audible-Alarm Dosimeters, U.S. Nuclear Regulatory Commission, August 1981.

17. Regulatory Guide 8.29, Instruction Concerning Risks from Occupational Radiation Exposure, Rev. 1, U.S. Nuclear Regulatory Commission, February 1996.

18. Regulatory Guide 8.34, Monitoring Criteria and Methods To Calculate Occupational Radiation Doses, U.S. Nuclear Regulatory Commission, July 1992.

19. Regulatory Guide 8.35, Planned Special Exposures, Rev. 1, U.S. Nuclear Regulatory Commission, August 2010.

12.1-13 Rev. 2

APR1400 DCD TIER 2

20. Regulatory Guide 8.36, Radiation Dose to the Embryo/Fetus, U.S. Nuclear Regulatory Commission, July 1992.

21. Regulatory Guide 8.38, Control of Access to High and Very High Radiation Areas of Nuclear Plants, Rev. 1, U.S. Nuclear Regulatory Commission, May 2006.

22. Regulatory Guide 4.21, Minimization of Contamination and Radioactive Waste Generation: Life-Cycle Planning, U.S. Nuclear Regulatory Commission, June 2008.

23. NEI 07-03A, Generic FSAR Template Guidance for Radiation Protection Program, Nuclear Energy Institute, May 2009.

24. NEI 08-08A, Generic FSAR Template Guidance for Life Cycle Minimization of Contamination, Nuclear Energy Institute, October 2009.

25. Regulatory Guide 8.4, Personal Monitoring Device-Direct Reading Pocket Dosimeters, Rev.1, U.S. Nuclear Regulatory Commission, June 2011.

12.1-14 Rev. 2

APR1400 DCD TIER 2 12.2 Radiation Sources This section identifies the sources of radiation within the plant that form the basis for shielding design calculations and the sources of airborne radioactivity used for the design of personnel protection measures and dose assessment.

12.2.1 Contained Sources The shielding design source terms are based on full-power operation with 0.25 percent fuel cladding defects and no credit for gas stripping. Sources in the primary coolant include fission products released from fuel cladding defects, activation products, and corrosion products. Throughout most of the RCS, activation products (principally nitrogen-16 [N-16]) are the primary radiation sources for shielding design during power operation.

12.2.1.1 Reactor Containment Building 12.2.1.1.1 Reactor Core The primary radiation emanating from the reactor core during normal operation consists of neutrons and gamma rays. Tables 12.2-1 and 12.2-2 list neutron and gamma fluxes in the reactor cavity at the side of the reactor vessel (RV), respectively. These tables are consistent with the parameters presented in Chapter 4. Table 12.2-3 lists gamma fluxes outside the RV after shutdown. The flux data represent the typical values, not the design values.

12.2.1.1.2 Reactor Coolant System Sources of radiation in the RCS are fission products released from the fuel and activation and corrosion products. The parameters used to calculate the reactor coolant fission product activities are listed in Table 12.2-4. The design basis reactor coolant activities are presented in Table 12.2-5. The reactor coolant fission product activities are calculated using Equations 11.1-1 and 11.1-2 (refer to Subsection 11.1.1.1).

Table 12.2-6 lists the average expected activities due to crud deposits on SG tubing and reactor system piping. The deposited crud activity is calculated using measured data from various operating PWRs and Equation 11.1-11 (refer to Subsection 11.1.3.1).

12.2-1 Rev. 2

APR1400 DCD TIER 2 The activation product, N-16, is the predominant activity source in the RCPs, SGs, and reactor coolant piping. The N-16 activity in each of the components depends on the transit time from the reactor core outlet to the component.

N-16 is produced by the O16(n, p)N16 reaction. The threshold energy for the reaction is 10.2 MeV. N-16 decays by beta emission with subsequent high-energy gamma emission 74 percent of the time. The gamma energies are 6.13 MeV, 69 percent of the time and 7.12 MeV, 5 percent of the time. The half-life of N-16 is 7.13 seconds.

The N-16 activity at the RV outlet nozzle is 5.53 x 106 disintegrations/cm3-sec. This activity, after considering the volume expansion of coolant at the vessel outlet nozzle, is calculated by the following expression and the reactor parameters. Table 12.2-7 lists typical N-16 activities at various locations in the primary coolant loop.

3 (1-e-tc )e-tr Activity (disintegrations/cm -sec) =

(1-e-tt )

Where:

= reaction rate (5.02 x 107 reactions/cm3-sec) tc = core transit time (0.81 sec) tt = total primary loop time (10.6 sec) tr = time from the active core outlet to the point of interest

= decay constant (0.097 sec-1)

The parameters and vapor activities in the pressurizer are listed in Table 12.2-27. The liquid activities in the pressurizer are assumed to be equal to the design basis reactor coolant activities in Table 12.2-5. The vapor activities of the pressurizer steam space are determined from the noble gases in Table 12.2-5. The vapor activities in the pressurizer are calculated using the following equation:

Vl mg Mwater xvsteam 2 ag = xm x xHXe or Kr xal Vg l RxT Where:

12.2-2 Rev. 2

APR1400 DCD TIER 2 ag, al = vapor and liquid activities, respectively, Bq Vg, Vl = vapor and liquid volume, respectively, cm3 mg, ml = vapor and liquid mass, respectively, gram Mwater = molecular weight of water, 18 gram/mol steam = specific volume of steam, cm3/gram R = ideal gas constant, 82.06 cm3-atm/°K-mol T = temperature, °K HXe or Kr = Henrys law constant for Xenon or Krypton, atm 12.2.1.1.3 Secondary-Side Systems The rate of SG tube leakage is assumed to be 3,270 L/day (0.6 gal/min). This is assumed to be concurrent with 0.25 percent fuel cladding defects for the shielding design basis source term calculations. A blowdown rate of 0.2 percent of maximum steaming rate is assumed. No credit is taken for the operation of the condensate polisher demineralizers.

Assumptions for calculation of main steam system (MSS) radiation sources are presented in Table 11.1-5, and results of the calculation are given in Table 12.2-18.

12.2.1.1.4 Spent Fuel Handling and Transfer The spent fuel assemblies are the predominant source of radiation in the reactor containment building (RCB) after plant shutdown for refueling. A reactor operating time to reach a near-equilibrium buildup level of fission products based on the rated power operation is used in determining the source strengths. The parameters used in the spent fuel decay gamma source calculation, such as fuel enrichment, specific core power, and discharge burnup, are given in Table 12.2-8. The spent fuel decay gamma source is given in Table 12.2-9. The neutron source term is not included because the dose rate contribution of neutron source term is negligible. Fuel assembly dose rates as a function of distance in the refueling pool water and time after shutdown are shown in Figures 12.2-1 and 12.2-2. The dose rates in the refueling pool are calculated by using neutron transport code, DORT (Reference 7), and flux-to-dose factors in ANSI/ANS-6.1.1 (Reference 8).

12.2-3 Rev. 2

APR1400 DCD TIER 2 These figures are based on the gamma source terms in Table 12.2-9. These figures provide the information of dose rates to an operator during refueling operation.

12.2.1.1.5 Processing Systems 12.2.1.1.5.1 Chemical and Volume Control System The shielding design for chemical and volume control system (CVCS) components is based on the maximum expected radioactivity in each of the components. The source term calculation is performed using SHIELD-APR (Reference 9) computer program.

Parameters of the source term calculation in CVCS are listed in Table 12.2-28. The component source terms are given in Tables 12.2-10 through 12.2-14.

a. Heat exchangers The radioactivity inventory for the CVCS heat exchanger is calculated using the following equations:

Qi ai AHX,i = (1-e-i tR ), for N-16 i

AHX,i =ai xVHX , for the other nuclides Where:

AHX,i = activity in heat exchanger, Bq Qi = influent flowrate, cm3/sec ai = influent specific activity, Bq/cm3 i = decay constant of a nuclide, sec-1 tR = transit time through heat exchanger, sec VHX = volume of heat exchanger, cm3 The maximum values for the CVCS heat exchanger (HX) radionuclide inventories are presented in Table 12.2-10.

12.2-4 Rev. 2

APR1400 DCD TIER 2 The total radioactivity inventory for the CVCS heat exchangers is based on the tube-side water volume (including the shell-side water volume for regenerative heat exchangers).

b. Ion exchangers The radioactivity inventory for the CVCS ion exchangers is calculated using the following equations:

Qi ai AIX,i = (1-e-it ) for DF1.0 i

AIX,i =ai xVIX,water for DF=1.0 Qi ai AIX,i = (1-e-itR ) for N-16 i

Where:

AIX,i = activity in ion exchanger, Bq Qi = influent flowrate, cm3/sec ai = influent specific activity, Bq/cm3

= removal efficiency of resin bed (=1-1/DF)

DF = decontamination factor i = decay constant of a nuclide, sec-1 t = replacement period of ion exchanger resin, sec tR = transit time through ion exchanger, sec VIX,water= water volume in ion exchanger, cm3 It is assumed that the porosity of resin is 0.5.

The maximum values for CVCS ion exchanger (IX) radionuclide inventories are presented in Table 12.2-11.

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APR1400 DCD TIER 2

1) Purification ion exchanger The total radioactivity inventory is based on resin buildup for the duration of 120 percent of one cycle of effective full-power days (EFPDs). This ion exchanger is used for removing lithium and purifying coolant water from RCS letdown. It is used for lithium removal for the first part of one-cycle EFPDs, which spans an average of 20 percent, and then it is put into service as a purification ion exchanger. The decontamination factor (DF) for anions is 100 with an efficiency of 99 percent. The DF for crud is 50 with an efficiency of 98 percent. The DF for all remaining nuclides except Xe, Kr, Y, Rb, and Cs is 50 with an efficiency of 98 percent. The DF for Xe, Kr, and Y is 1 with an efficiency of 0 percent. The DF for Rb and Cs is 2 with an efficiency of 50 percent. The radioactivity inventory in processed liquid is based on a normal letdown flow rate.

2) Deborating ion exchanger The total radioactivity inventory is based on resin buildup during the end-of-cycle (EOC) deborating operation. This ion exchanger is used to reduce reactor coolant boron concentration at the EOC. Boron control in the CVCS is described in detail in Subsection 9.3.4. The DF for nuclides except anions and crud is 1 with an efficiency of 0 percent. The DF for anions and cruds is 10 with an efficiency of 90 percent. The radioactivity inventory in processed liquid is based on a normal letdown flow rate.

3) Pre-holdup ion exchanger The total radioactivity inventory is based on resin buildup for the duration of one-cycle EFPDs. The DF for all nuclides except Xe, Kr, Y, Rb, and Cs is 10 with an efficiency of 90 percent. The DF for Rb and Cs is 100 with an efficiency of 99 percent; a 3-to-1 cation to anion ratio provides a minimum DF of 100.. The DF for Xe, Kr, and Y is 1 with an efficiency of 0 percent.

Liquid inputs processed by the pre-holdup ion exchanger include the letdown processed through the purification ion exchanger and the purification filter, and the inflow from the reactor drain tank (RDT) and the equipment drain tank (EDT).

4) Boric acid condensate ion exchanger 12.2-6 Rev. 2

APR1400 DCD TIER 2 The total radioactivity inventory is based on resin buildup for the duration of one-cycle EFPDs. The DF for anions is 10 with an efficiency of 90 percent.

The DF for other nuclides including crud is 1 with an efficiency of 0 percent.

The radioactivity inventory is based on input from the letdown processed through the purification ion exchanger and the purification filter, the inflow from the RDT and the EDT, and the inflow from the boric acid concentrator (BAC) in consideration of the average condensation rate.

c. Filters The radioactivity inventory for the CVCS filters is calculated using the following equations:

Qi ai AFL,i = (1-e-it ) for DF1.0 i

AFL,i =ai xVFL for DF=1.0 Qi ai AFL,i = (1-e-itR ), for N-16 i

Where:

AFL,i = activity in filter element, Bq Qi = influent flowrate, cm3/sec ai = influent specific activity, Bq/cm3

= removal efficiency of filter element (=1-1/DF)

DF = decontamination factor of filter element i = decay constant of a nuclide, sec-1 t = replacement period of filter element, sec tR = transit time through filter, sec VFL = filter volume, cm3 12.2-7 Rev. 2

APR1400 DCD TIER 2 The maximum values for CVCS filter radionuclide inventories are presented in Table 12.2-12.

The total radioactivity inventories in the CVCS filters are based on crud buildup for the duration of one-cycle EFPDs. The DF assumed for crud buildup in the filter is 10 with an efficiency of 90 percent.

d. Tanks The radioactivity inventory for the CVCS tanks is calculated using the following equations:

For Liquid activity, Q

Qi ai -i + i t AL,i = Q (1-e VL

)

i + i VL For Noble Gases at equilibrium condition, (18)(H)(VV )(AL)

AV = (VL )(R)(T)

For Noble Gases at non-equilibrium condition, PF Qi ai AV,i =( PF+1 )( )(1-e-it )

i For Particulates, PF AV =( PF+1 )AL For new liquid activity (particulates and noble gases at non-equilibrium condition) 1

= (+1)

Where:

AL, AV = Liquid and Vapor activity, respectively, Bq VL, VV = Liquid and Vapor volume, respectively, cm3 12.2-8 Rev. 2

APR1400 DCD TIER 2 Qi = influent flowrate, cm3/sec ai = influent specific activity, Bq/ cm3 i = decay constant of a nuclide, sec-1 PF = partition factor t = transit time through tank, sec 18 = molecular weight of water, gram/mol H = Henrys constant, atm R = ideal gas constant, 82.06 cm3-atm/°K-mol T = temperature, °K The liquid and vapor volumes are obtained from the low water level to maximize the VCT vapor source terms. The liquid source terms in VCT are additionally considered for the water volume from low water level to high water level. This maximizes the vapor and the liquid source terms in VCT.

The activity of the vapor space is determined initially by using Henrys law. For non-equilibrium condition, partition factors are used to re-evaluate the vapor space activity. In the case of particulates and non-equilibrium noble gases, a new liquid activity is calculated after accounting for partitioning.

The maximum values for CVCS tank radionuclide inventories are presented in Table 12.2-13.

The total radioactivity inventories in the CVCS tanks are based on buildup for the duration of one-cycle EFPDs.

e. Boric acid concentrator The radioactivity inventory for the CVCS boric acid concentrator is calculated using the following equations. The influent specific activities of concentrate heater are assumed to be equal to the specific activities of holdup pump.

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APR1400 DCD TIER 2

1) Concentrate Heater 1

aheater,i =ai ( 1+PF ) for noble gas aheater,i =ai xCFBAC , for all nuclides except noble gas and N-16 Aheater,i =aheater,i xVheater Where:

aheater,i = specific activity of concentrate heater, Bq/cm3 PF = partition factor for BAC heater (=10)

CFBAC = BAC concentration factor (=100)

Vheater = BAC heater volume, cm3 Aheater,i = concentrate heater activity, Bq

2) Concentrate Cooler Acooler,i =aheater,i xVcooler Where:

Vcooler = concentrate cooler volume, cm3 Acooler,i = concentrate heater activity, Bq

3) Flash Tank The influent vapor specific activities of flash tank are assumed to be equal to the specific activities of holdup pump. The specific activities of BAC heater used to the influent liquid specific activities of BAC flash tank.

Aflash tank,i =ai xVV,flash tank for noble gas Aflash tank,i =aheater,i xVL,flash tank for all nuclides except noble gas Where:

12.2-10 Rev. 2

APR1400 DCD TIER 2 VV flash tank, VL flash tank = vapor and liquid volume in flash tank, resp ectively, cm3 Aflash tank,I = flash tank activity, Bq

4) Vapor Separator PF aseparator,i =ai 1+PF for noble gas aseparator,i =ai for all nuclides except noble gas Aseparator,i =aseparator,i xVV,separator for noble gas Aseparator,i =aseparator,i xVL,separator for all nuclides except noble gas Where:

aseparator,i = specific activity of vapor separator, Bq/cm3 PF = partition factor for BAC vapor separator

(=10)

VV separator, VL separator = vapor and liquid volume in vapor separator, respectively, cm3 Aseparator,i = vapor separator activity, Bq

5) Concentrate Pump, Aconc pump,i =aheater,i xVconc pump Where:

Vconc pump = concentrate pump volume, cm3 Aconc pump,i = concentrate pump activity, Bq

6) Concentrate Transfer Pump Atrans pump,i =aheater,i xVtrans pump 12.2-11 Rev. 2

APR1400 DCD TIER 2 Where:

Vtrans pump, = concentrate transfer pump volume, cm3 Atrans pump,i = concentrate transfer pump activity, Bq The maximum values for BAC radionuclide inventories are presented in Table 12.2-14. The components of the boric acid concentrator (BAC) are conservatively modeled with the combined total source term based on the maximum design basis source terms in order to determine the minimum required shield wall thicknesses for the BAC room to maintain the dose rates in the adjacent rooms and areas below their respective radiation zone limits. The shielding analyses are performed for the vapor phase and liquid phase for the BAC components using their corresponding source terms as listed in Table 12.2-14.

The dose rate results calculated for the two phases are summed for the determination of the minimum shield wall thicknesses and corresponding radiation zone for the BAC room.

The total radioactivity inventories in the BAC package are based on a concentration factor of 100.

12.2.1.1.5.2 Steam Generator Blowdown System Radiation sources in the steam generator blowdown system (SGBS) are shown in Table 12.2-18. The sources are based on the assumed design basis primary-to-secondary (PTS) leakage rate and the assumed fuel defect percentage described in Subsection 12.2.1.1.3.

The blowdown rate is assumed to be 1 percent of the maximum steaming rate. The nuclide accumulation in the SGBS pre-filter, post-filters, and the mixed beds are calculated based on radioactive crud and nuclide buildup at the end of 6 months of processing.

12.2.1.1.5.3 Condensate Polishing System Radiation sources in the condensate polishing system (CPS) are shown in Table 12.2-18.

The sources are based on the design basis PTS leakage and the assumed fuel defect percentage described in Subsection 12.2.1.1.3. It is assumed that 65 percent of the total steam flow rate and the SGBS blowdown flow through the CPS and that one pair (16-2/3 percent of the condensate flow) out of total seven pairs of CPS demineralizers, of which one pair of demineralizers is in standby service, is used to process the condensate during 12.2-12 Rev. 2

APR1400 DCD TIER 2 normal operation. The nuclide accumulation in the CPS cation bed and the mixed bed are based on radioactivity buildup at the end of their corresponding processing cycle. The nuclide accumulation times for the cation bed and the mixed bed ion exchangers are about 3 days and 30 days, respectively. The cation and mixed bed ion exchange resins are then replaced at the end of the processing times accordingly.

12.2.1.1.6 Gamma Sources of Irradiated Components The components in the reactor vessel are irradiated by the fission neutrons during the core power operation and are activated. The in-core instrument (ICI) assembly, which consists of five rhodium detectors, one background detector, one core-exit thermocouple, and a central member assembly, is enclosed in a protective sheath. Activated gamma sources of the irradiated ICI assembly are estimated assuming 6 years of irradiation. The activated gamma sources of the irradiated control element assembly (CEA) and the irradiated neutron source assembly (NSA) are estimated assuming 10 years of irradiation. In CEA, the neutron absorbing material is B4C and the cladding material is Inconel 625. The NSA contains the primary neutron source of Cf252 and the secondary neutron source of Sb-Be.

The activated gamma source of the irradiated surveillance capsule assembly (SCA) is estimated assuming 60 years of irradiation. The irradiation calculation is performed using the ORIGEN-S computer program, and the irradiation time of each component is based on the design life time of each component. The activated gamma sources of ICI, CEA, NSA, and SCA are given in Table 12.2-15.

The handling and storage of the irradiated ICI and CEAs during refueling operation are described in Subsection 9.1.4.

Dimensions and parameters of the other radiation sources in reactor containment building used in the shielding analyses are listed in Table 12.2-25.

12.2.1.2 Auxiliary Building 12.2.1.2.1 Shutdown Cooling System The shutdown cooling system (SCS) pumps, heat exchangers, and associated piping are considered to be contaminated with radioactive materials. During plant shutdown, radioactive isotopes carried in the reactor coolant after 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of decay following shutdown are transported into the SCS pumps and heat exchangers. The maximum radioactive source strengths in the SCS are provided in Table 12.2-16.

12.2-13 Rev. 2

APR1400 DCD TIER 2 12.2.1.2.2 Component Cooling Water System The component cooling water system (CCWS) is a closed-loop demineralized water system that can potentially become contaminated by heat exchanger leakage from other radioactive systems that are cooled by the CCWS. The CCWS is provided with radiation monitors to detect inleakage of radioactive contaminants as described in Subsection 11.5.2.3. The detection of leakage is sensitive to a level of 3.7 x 10-2 Bq/cm3 gross gamma activity in the CCWS cooling water.

In the event of a major leak in a CCWS train, the train is removed from service, a nonleaking train is used, and the contaminated water is treated through the liquid waste management system (LWMS). This design prevents the spread of contamination through the CCWS and connecting systems.

12.2.1.2.3 Spent Fuel Storage and Transfer Spent fuel assemblies and associated crud are the primary sources of radiation in the spent fuel storage and transfer area. The shielding design assumes the maximum number of spent fuel assemblies in storage.

12.2.1.2.4 Spent Fuel Pool Cooling and Cleanup System Activity levels in the spent fuel pool cooling and cleanup system (SFPCCS) are determined based on the activities present in the spent fuel pool (SFP). Design basis SFP activities are presented in Table 12.2-17.

SFP fission and corrosion product specific activities are evaluated in the initial stage of refueling. It is assumed that the reactor coolant is cooled down for approximately 2 days during a shutdown for refueling. During this period, the primary coolant is let down through the purification filter, purification ion exchanger, gas stripper, and volume control tank (VCT).

The purification process serves two purposes:

a. Removing the noble gases in the gas stripper avoids large activity releases to the reactor containment building following reactor vessel head removal.

12.2-14 Rev. 2

APR1400 DCD TIER 2

b. The ion exchange and filtration reduces dissolved fission and corrosion product concentrations in the coolant that would otherwise enter the SFP and the refueling pool.

At the end of the 2-day period, the coolant above the reactor vessel flange is partially drained. The RV head is unbolted, and the refueling pool is filled with the water from the in-containment refueling water storage tank (IRWST). The remaining reactor coolant volume is then mixed with water in the refueling pool and the SFP.

After refueling, the SFP is isolated and the water in the refueling pool is returned to the IRWST. This series of events determines the total activity in the SFP. The fission and corrosion product activities in the SFP, listed in Table 12.2-17, are calculated using the reactor coolant equilibrium concentrations presented in Table 12.2-5. The SFP activities are subsequently reduced by decay during refueling as well as by operation of the SFPCCS.

The source terms for the SFP demineralizers and filters are provided in Table 12.2-17a.

The activities are integrated over the cleanup time for normal (expected) operation and are determined to be at maximum at about 265 hours0.00307 days <br />0.0736 hours <br />4.381614e-4 weeks <br />1.008325e-4 months <br /> and 290 hours0.00336 days <br />0.0806 hours <br />4.794974e-4 weeks <br />1.10345e-4 months <br /> for the SFP filter and demineralizer respectively, after which time the source terms decrease due to decay of short half-life nuclides.

There is no contribution from defective fuel elements because of low power and temperature during plant shutdown operations.

Dimensions and parameters of the radiation sources in auxiliary building used in the shielding analyses are listed in Table 12.2-25.

12.2.1.3 Turbine Generator Building Radiation sources in the turbine generator building occur in the condensate polishing system (CPS) due to the design basis PTS leakage rate in the steam generator. Activity levels for all turbine generator building related sources are summarized in Table 12.2-18.

The activities provided in Table 12.2-18 are based on normal operation reactor coolant activity levels and PTS leakage conditions. Radionuclide removal efficiencies of demineralizers in the CPS are assumed to be consistent with the guidance in NUREG-0017 (Reference 1).

12.2.1.4 Compound Building 12.2-15 Rev. 2

APR1400 DCD TIER 2 Radioactive sources in the radwaste system components include fission and activation radionuclides produced in the core and in the reactor coolant. The level of radioactivity is dependent on the components and operating parameters of the particular radwaste system.

Gaseous radwaste system (GRS) source terms are provided in Table 12.2-19. Radiation sources for each component of the GRS are calculated using the shielding basis equilibrium reactor coolant radionuclide concentrations provided in Table 12.2-5, which are based on an assumed 0.25 percent fuel defect. Activity buildup on the process gas charcoal beds is calculated assuming maximum design basis holdup times for noble gases in accordance with NUREG-0017.

The source terms for LWMS tanks are provided in Table 12.2-20 and for the other LWMS processing equipment in Table 12.2-21. Source terms for the equipment waste tank (EWT) and floor drain tank (FDT) are calculated using reactor coolant equilibrium radionuclide concentrations presented in Table 12.2-5 and the activity fractions in Table 11.2-2.

Radionuclide concentrations in the LWMS are determined using the DIJESTER Code (Reference 2). The accumulation and decay of radionuclides in the LWMS can be modeled using this code.

The activities of LWMS demineralizers are calculated using 44% of the RCS source term and an activity buildup and decay model. The calculation applies the process flow rates provided in Table 11.2-2, and the process fluid activity levels provided in Table 12.2-20.

Source terms for the monitor tank are calculated using reactor coolant equilibrium radionuclide concentrations with 0.25% failed fuel and processed through the LWMS.

The demineralizer resin is assumed to have a service life of 1 year. Although the service life of filters and resins in the LWMS may vary according to operating conditions, for radiation protection purposes, they are replaced based on the source term strength to provide reasonable assurance that occupational exposures associated with radwaste system operations remain ALARA.

Solid waste management system (SWMS) source terms are provided in Table 12.2-22.

Source terms for the spent resin long-term storage tank are calculated based on 10 years of cumulative radioactive resin batches from the CVCS demineralizers with decay. Source terms for the LWMS demineralizers are presented in Table 12.2-21, based on a processing time of 1 year. The source terms for the low-activity spent resin tank (LASRT) in Table 12.2-22 are calculated using the source term for the LWMS spent resin and multiplying by a factor of 3 for conservatism in the shielding design. Since the LASRT is filled with 12.2-16 Rev. 2

APR1400 DCD TIER 2 resins from the LWMS, SGBD, and SFPCCS for 1 year and the source terms for the LWMS are highest, it is conservative that the LASRT is only filled with LWMS resin.

Additionally, the source term in the waste drum storage area is presented in Table 12.2-22a.

Dimensions and parameters of the radiation sources in compound building used in the shielding analyses are listed in Table 12.2-25.

12.2.1.5 Sources Resulting from Design Basis Accidents Design parameters and source terms for design basis accidents (DBAs) are addressed in Chapter 15.

12.2.1.6 Stored Radioactivity The holdup tank, reactor makeup water tank (RMWT), and boric acid storage tank (BAST) are the principal sources of activity outside the plant buildings. The surface dose rate of these tanks is designed so that it does not exceed 2.5 Sv/hr, except the immediate areas outside the side manways for these outdoor tanks which may exceed the Zone 1 criterion.

A physical barrier and administrative controls are in place to prevent personnel from occupying the immediate vicinity of the outside tanks.

Spent fuel is stored in the SFP until it is placed in the spent fuel shipping cask for transport to an onsite interim storage facility or to an offsite storage facility.

Storage space is allocated in the compound building for the storage of spent filter cartridges and dewatered spent resin in the spent filter drum and HIC storage area as well as for solidified R/O concentrates and dry active waste (DAW) in drums in the waste drum storage area. The shielding design for the spent filter drum and HIC storage area is based on using the expected stored waste volumes for normal operation and the design basis source term (0.25% fuel failure) for the activity of wastes. For the source term of HICs, the volume and the associated source term of spent resin (not decayed) is increased by a factor of 1.656 (=Volume of 16 HICs / 1-cycle volume of spent resin) for conservatism.

The zoning for this area is determined by summing the dose rates from the HIC and the spent filter drums. However, in determining the minimum shield wall thicknesses, the two dose rates are calculated individually since the impact of shield wall thicknesses is dominated by the close proximity of the individual sources (HIC or spent filter drum) to the walls around the designated storage areas. The shielding design for the waste drum area is based on the use of the design basis source term for the solidified R/O concentrate and the 12.2-17 Rev. 2

APR1400 DCD TIER 2 source term for the DAW is based on a waste drum with the highest activity at Korean domestic nuclear power plants to ensure that the shielding design is sufficiently conservative. Some of the drummed waste, such as R/O concentrate, are expected to have higher activity, and are stored at the interior of the drum layers within a stacking configuration, with lower activity waste drums at the exterior of each layer, to provide additional shielding.

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 12.2(1)).

12.2.1.7 Pipe Routing Piping carrying radioactive materials is routed in pipe chases to the extent practicable when routed through low-radiation and low-contamination areas to maintain radiation exposure to plant personnel ALARA and reduce the spread of contamination.

Criteria for routing radioactive piping include:

a. Piping containing radioactive material is routed through shielded pipe chases to the extent practicable.

b. Systems containing radioactive liquids, gases, or slurries are physically located close to interfacing systems to reduce pipe length and minimize the need for routing radioactive piping through personnel access corridors.

c. Stagnant runs of piping are avoided to minimize the potential for crud traps.

Flushing and decontamination capabilities are provided as necessary.

12.2.2 Airborne Radioactive Material Sources Airborne radioactive material is introduced into the plant atmosphere through the leakage of radioactive fluids from equipment (e.g., valve stems, pump seals) and the evaporation of tritiated water.

The APR1400 is designed to minimize the potential for leakage of radioactive fluids.

However, leakage of radioactive fluids through pump seals, valve stems, and other piping connections may not be eliminated entirely. According to the recommendations and guidance provided by NUREG-0017, airborne source terms are estimated based on the type 12.2-18 Rev. 2

APR1400 DCD TIER 2 of equipment, number of valves, pumps, and flanges, and level of radioactivity in the fluid stream.

The evaporation of tritiated water from unenclosed sources of water, such as the SFP, is a significant source of airborne tritium. The rate of evaporation is dependent on the pool temperature, air velocity across the pool, and relative humidity.

The airborne radioactivity concentrations in rooms or cubicles of occupied plant areas are maintained within the in-plant concentration limits prescribed in the applicable regulations.

The plant design criteria that are implemented to provide reasonable assurance of conformance with the applicable regulations are as follows:

a. Maintenance of in-plant airborne concentrations of radioisotopes in normally occupied areas within the derived air concentration (DAC) prescribed in Table 1 of 10 CFR Part 20, Appendix B (Reference 3).

b. Sufficient confinement and ventilation capability to prevent the spread of airborne contamination.

c. Airborne radiation monitors in normally occupied areas where the potential for airborne contamination exists. The airborne radiation monitoring system is described in Section 11.5.

12.2.2.1 Production of Airborne Radioactive Material Radioactive materials become airborne through the natural evaporation of contaminated fluids. Major contributors to airborne radioactivity during normal operation are (1) CVCS leaks, (2) evaporation from the SFP, (3) radwaste system leaks, (4) venting of radwaste tanks, and (5) HVAC system leaks. Minor contributions are from (1) cleaning and decontamination tools and equipment, (2) contaminated clothing, and (3) sample preparation and analysis.

Abnormal occurrences that can cause airborne radioactivity include (1) spills such as overflows and splashing, (2) failure of a ventilation system, (3) cracks and breaks in piping, (4) failures of pump and valve seals, and (5) equipment malfunctions.

12.2.2.2 Airborne Sources in Normally Accessible Areas 12.2-19 Rev. 2

APR1400 DCD TIER 2 Airborne radioactive material is expected to affect general access areas and may become worse in the event of ventilation system failure or spillage of radioactive material in areas that are not sealed from general access areas. The ventilation pathway is designed to flow from areas of low airborne radioactivity to areas of potentially greater airborne radioactivity.

The ventilation system is designed to control the airborne radioactivity in the laboratories, maintenance areas, and the reactor containment building.

Radiation exposure due to maintenance activities accounts for the majority of the exposure received by plant personnel because many maintenance activities are performed in areas with relatively high radiation and airborne radioactivity. The airborne radionuclide concentrations are calculated for the most common contributors, such as leaks and venting.

12.2.2.3 Airborne Concentrations The determination of concentrations of airborne radioactive nuclides in cubicles is based on the model described below.

Concentrations of noble gases, tritium, and iodine are calculated as a fraction of the DAC for air in the reactor containment building and normally accessible cubicles within the auxiliary building and the compound building. Except under abnormal conditions, general access areas normally have little, if any, airborne contaminants during normal operation.

Table 12.2-23 provides the airborne radioactivity concentrations and DAC fractions for the main cubicles in the reactor containment, auxiliary building, and compound building.

Equilibrium airborne radioactivity concentrations in rooms, cubicles, and other areas during normal operation are calculated based on the following equation:

LCP CA =

(V+F)

Where:

CA = airborne concentration in each cubicle (Bq/cm3)

L = leak rate (cm3/min)

C = radioactive concentration of liquid or gas (Bq/cm3)

P = fraction of activity released to air 12.2-20 Rev. 2

APR1400 DCD TIER 2

= decay constant (min-1)

V = enclosed room volume (cm3)

F = air exhaust flow rate (cm3/min)

The equation above represents the solution of a differential equation in equilibrium conditions in which there is no airborne radioactivity in the ventilation airflow(s) entering the area under consideration. To accommodate situations in which airborne radioactivity is in one or more ventilation airflow streams entering the area of concern, additional term(s) are added to the basic differential equation. Minimum HVAC flow rates and parameters used in the airborne source term calculations are provided in Table 12.2-26.

12.2.2.4 Basis and Assumptions for Partition Factors

a. Reactor Containment Building The partition factors for airborne activity in the reactor containment building are calculated as follows:

1) 1.0 for noble gases

2) All other nuclides: use flashing fraction (ff), based on the enthalpy difference in accordance with RG 1.183.

b. Auxiliary Building and Compound Building The partition factors for airborne activity in the auxiliary building are calculated as follows:

1) 1.0 for noble gases

2) 1 x 10-3 for halogens in cold liquid (<120 °F) and 0.1 for halogens in a hot liquid (>120 °F)

3) 0.53 for H3 in primary coolant and 0.1 for H3 in cold liquids, and

4) 0.005 for all other nuclides 12.2-21 Rev. 2

APR1400 DCD TIER 2 12.2.3 Sources Used in NUREG-0737 Post-Accident Shielding Analysis Item II.B.2.3 of NUREG-0737 (Reference 4) clarifies the requirement for providing reasonable assurance that areas that require post-accident personnel access or contain safety-related equipment are adequately shielded in the vicinity of systems that may contain highly radioactive materials as a result of a DBA.

Plant areas requiring post-accident occupation (vital areas), their locations and the expected doses are identified in Subsections 12.3.1.8, 12.3.1.9 and 12.4.1.2, respectively.

The review of the shielding for systems that, as a result of an accident, contain highly radioactive materials is performed using the methodology described in Subsection 12.3.2.

Initial core releases that are equivalent to those recommended in NRC RG 1.183 (Reference

5) are used. The source terms are presented in Table 12.2-24.

These source terms are used to evaluate the adequacy of shielding in post-accident conditions using shielding computer codes, which are addressed in Section 12.3.2.2, to verify that:

a. Vital areas are accessible for operators to take the required mitigative actions during post-accident conditions.

b. Safety-related equipment is qualified for the environmental radiation condition in the area where the equipment is located.

In vital areas that are not occupied continuously, sufficient shielding is provided for reasonable assurance that radiation exposure during post-accident conditions does not exceed the total effective dose equivalent (TEDE) criterion (50 mSv). In vital areas that are occupied continuously, such as the main control room (MCR), the local radiation levels are limited to 0.15 mSv/hr averaged over 30 days per NUREG-0737, Subsection II.B.2.3 (Reference 4).

12.2.4 Combined License Information 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.

12.2-22 Rev. 2

APR1400 DCD TIER 2 12.2.5 References

1. NUREG-0017, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors, Rev. 1, U.S. Nuclear Regulatory Commission, April 1985.

2. DIJESTER, A Program to Compute Radioactive Decay in Fluid Flow Systems, Program No. 9.8. 060-1.0, D. J. Pichurski, April 1976.

3. 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.

4. NUREG-0737, Item II.B.2.3, Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces/Systems Which May Be Used in Post-Accident Operations, U.S. Nuclear Regulatory Commission, 1980.

5. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, U.S. Nuclear Regulatory Commission, July 2000.

6. ANSI/ANS-18.1-1999, Radioactive Source Term for Normal Operation of Light Water Reactors, American Nuclear Society, 1999.

7. CCC-650/DOORS3.2, One-, Two- and Three- Dimensional Discrete Ordinates Neutron/Photon Transport Code System, Radiation Safety Information Computational Center, ORNL, 1998.

8. ANSI/ANS-6.1.1, American National Standard for Neutron and Gamma-Ray Fluence-to-Dose Factors, American Nuclear Society, August 1991.

9. 00000-CV-VV-002, Rev.00, Software Verification and Validation Report for SHIELDAPR, KEPCO E&C, October 2002.

12.2-23 Rev. 2

APR1400 DCD TIER 2 Table 12.2-1 Normal Operation Neutron Spectra Outside the Reactor Vessel Average Neutron Neutron Spectra(1) Average Neutron Neutron Spectra(1)

Energy (eV) (neutrons/cm2-s) Energy (eV) (neutrons/cm2-s) 1.58E+07 3.84E+05 2.40E+05 1.88E+09 1.32E+07 1.04E+06 1.47E+05 1.90E+09 1.11E+07 3.78E+06 8.92E+04 1.23E+09 9.30E+06 6.69E+06 5.41E+04 9.14E+08 8.01E+06 9.76E+06 3.64E+04 3.04E+08 6.74E+06 2.03E+07 2.89E+04 1.82E+08 5.52E+06 2.79E+07 2.51E+04 5.46E+08 4.32E+06 5.00E+07 2.30E+04 3.50E+08 3.35E+06 4.00E+07 1.85E+04 5.96E+08 2.87E+06 3.17E+07 1.11E+04 7.74E+08 2.60E+06 3.94E+07 5.23E+03 7.96E+08 2.42E+06 2.00E+07 2.47E+03 6.63E+08 2.36E+06 6.20E+06 1.02E+03 1.03E+09 2.29E+06 3.13E+07 3.34E+02 5.40E+08 2.08E+06 8.73E+07 1.58E+02 5.33E+08 1.79E+06 1.22E+08 6.93E+01 6.54E+08 1.50E+06 2.07E+08 2.40E+01 7.57E+08 1.18E+06 4.96E+08 7.86E+00 4.19E+08 9.12E+05 5.06E+08 3.45E+00 5.34E+08 7.82E+05 2.31E+08 1.37E+00 3.89E+08 6.75E+05 1.04E+09 6.45E-01 3.55E+08 5.53E+05 1.01E+09 2.57E-01 7.95E+08 4.33E+05 9.84E+08 5.00E-02 3.06E+09 3.33E+05 1.53E+09 (1) At core midplane, 15.24 cm (0.5 ft) from vessel surface 12.2-24 Rev. 2

APR1400 DCD TIER 2 Table 12.2-2 Normal Operation Gamma Spectra Outside the Reactor Vessel Average Gamma Gamma Spectra (1)

Energy (eV) (Gamma/cm2-s) 1.20E+07 3.75E+05 9.00E+06 1.29E+08 7.50E+06 5.21E+08 6.50E+06 3.45E+08 5.50E+06 2.79E+08 4.50E+06 4.19E+08 3.50E+06 5.14E+08 2.50E+06 9.83E+08 1.75E+06 7.36E+08 1.25E+06 7.71E+08 9.00E+05 5.04E+08 7.50E+05 2.48E+08 6.50E+05 2.97E+08 5.00E+05 1.32E+09 3.00E+05 2.08E+09 1.50E+05 1.81E+09 8.00E+04 4.80E+08 4.50E+04 3.05E+07 2.50E+04 3.45E+05 1.50E+04 1.42E+05 (1) At core midplane, 15.24 cm (0.5 ft) from vessel surface 12.2-25 Rev. 2

APR1400 DCD TIER 2 Table 12.2-3 Shutdown Gamma Spectra Outside the Reactor Vessel Average Gamma Decay Gamma(1) Material Activation Energy (eV) (Gamma/cm2-sec) (Gamma/cm2-sec) 4.50E+06 0.00E+00 4.20E+02 3.50E+06 9.11E+02 4.45E+01 2.50E+06 9.70E+04 8.59E+05 1.75E+06 1.95E+05 1.33E+05 1.25E+06 3.03E+05 9.02E+06 9.00E+05 1.67E+05 1.60E+06 7.50E+05 9.73E+04 6.65E+05 6.50E+05 1.11E+05 7.38E+05 5.00E+05 2.89E+05 2.06E+06 3.00E+05 5.03E+05 4.98E+06 1.50E+05 4.21E+05 3.99E+06 8.00E+04 1.05E+05 1.03E+06 4.50E+04 5.25E+03 5.31E+04 2.50E+04 1.09E+00 1.10E+01 1.50E+04 0.00E+00 0.00E+00 (1) At core midplane, 15.24 cm (0.5 ft) from vessel surface, and 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shutdown 12.2-26 Rev. 2

APR1400 DCD TIER 2 Table 12.2-4 Basis for Reactor Coolant Fission Product Activities(1)

Parameter Maximum Core power level, MWt 4,063 Duration of reactor operation, fuel cycles 5 Equilibrium fuel cycle, equivalent full-power days 480 2

Thermal neutron flux, n/cm -sec 6.32E+13 Average thermal fission rate, fission/MW-sec 3.12E+16 Fraction of fuel defect 0.0025 Reactor coolant mass including pressurizer, kg (lb) 2.92E+05 (6.43E+05)

Core coolant volume to reactor coolant volume ratio 0.073 Purification flow, kg/sec (lb/sec) 5.02 (11.07)

Boron concentration reduction rate, ppm/second 2.60E-05 Fuel cycle boron concentration at BOC (minimum), ppm 1,110 Ion exchanger and gas stripper removal efficiency CVCS purification ion exchanger Noble gas, tritium 0.0 Cs, Rb 0.5 Anion 0.99 All others 0.98 CVCS gas stripper removal efficiency Noble gas 0.999 All others 0.0 CVCS gas stripper operation None Fission product escape rate coefficients (sec-1)

Xe, Kr 6.5E-08 I, Br, Rb, Cs 1.3E-08 Mo 2.0E-09 Te 1.0E-09 Sr, Ba 1.0E-11 All others 1.6E-12 (1) Shielding design source term (0.25% fuel defect, DAMSAM code input) 12.2-27 Rev. 2

APR1400 DCD TIER 2 Table 12.2-5 (1 of 2)

Reactor Coolant Equilibrium Concentration (Core Power: 4,063 MWt, 0.25 % Fuel Defect, No Gas Stripping)

Nuclide Activity (Bq/g) Nuclide Activity (Bq/g)

Kr-85m 1.04E+04 Co-60 2.10E+01(1)

Kr-85 4.44E+04 Zn-65 2.02E+01(1)

Kr-87 8.14E+03 Sr-89 3.26E+01 Kr-88 2.26E+04 Sr-90 2.18E+00 Xe-131m 4.44E+04 Sr-91 4.81E+01 Xe-133m 2.70E+03 Y-91m 2.81E+01 Xe-133 2.89E+06 Y-91 4.81E+00 Xe-135m 5.92E+03 Y-93 1.15E+00 Xe-135 5.92E+04 Zr-95 1.54E+01(1)

Xe-137 1.37E+03 Nb-95 5.18E+00 Xe-138 5.18E+03 Mo-99 2.81E+03 Br-84 1.96E+02 Tc-99m 1.63E+03 I-131 2.48E+04 Ru-103 1.74E+00 I-132 6.66E+03 Ru-106 7.40E-01 I-133 3.52E+04 Ag-110m 5.15E+01(1)

I-134 4.07E+03 Te-129m 5.92E+01 I-135 2.00E+04 Te-129 6.29E+01 Rb-88 2.29E+04 Te-131m 2.81E+02 Cs-134 3.52E+03 Te-131 1.11E+02 Cs-136 4.81E+02 Te-132 1.96E+03 Cs-137 4.07E+03 Ba-137m 3.70E+03 (2)

N-16 8.22E+06 Ba-140 4.07E+01 H-3 1.30E+05 La-140 1.37E+01 (1)

Na-24 1.81E+03 Ce-141 1.52E+00 12.2-28 Rev. 2

APR1400 DCD TIER 2 Table 12.2-5 (2 of 2)

Nuclide Activity (Bq/g) Nuclide Activity (Bq/g)

Cr-51 5.48E+02 Ce-143 4.07E+00 (1)

Mn-54 6.34E+01 Ce-144 4.44E+00 (1)

Fe-55 4.75E+01 W-187 9.70E+01(1)

Fe-59 1.19E+01(1) Np-239 8.62E+01(1)

Co-58 1.82E+02(1)

(1) Expected source terms based on ANSI/ANS 18.1 (Reference 6) are used when these values are higher than the design basis source terms, for added conservatism.

(2) At the reactor vessel outlet nozzle 12.2-29 Rev. 2

APR1400 DCD TIER 2 Table 12.2-6 Deposited Crud Activities (SG Tubing and Reactor System Piping)

Activity Isotope (Bq/g-crud)

Cr-51 7.31E+09 Mn-54 1.99E+07 Fe-59 4.18E+07 Co-58 1.77E+09 Co-60 6.22E+07 Zr-95 4.90E+07 12.2-30 Rev. 2

APR1400 DCD TIER 2 Table 12.2-7 N-16 Activity Accumulated Location Activity (Bq/g) Transit Time (sec)

Vessel outlet nozzle 8.22E+06 0.66 Vessel outlet line (midpoint) 8.13E+06 0.78 Steam generator (midpoint) 5.68E+06 3.85 Reactor coolant pump (midpoint) 3.78E+06 7.56 Vessel inlet line (midpoint) 3.61E+06 8.02 Containment penetration area 1.94E+04 62.43(1)

(1) The N-16 concentration at the containment penetration is estimated to be 1.94E+04 Bq/g based on the transit time from RV to SG and the residence times in the regenerative and the letdown heat exchangers before the containment penetration. For shielding analysis purposes, this concentration is assumed to be the same as that at the outlet of the letdown heat exchanger, and no credit is taken for further decay along the 200 ft of piping from the letdown heat exchanger outlet to the containment isolation valve.

12.2-31 Rev. 2

APR1400 DCD TIER 2 Table 12.2-8 Parameters Used in Spent Fuel Decay Gamma Source Calculation Parameter Value Core power level, MWt 4,063 No. of fuel assemblies in core 241 Uranium loading, kg U/assembly 430.8 235 Fuel initial enrichment, U w/o 4.20 Irradiation cycle length, EFPD 480 Burnup per cycle, GWD/MTU 18.8 No. of irradiation cycles 3 Discharge burnup, GWD/MTU 56.4 12.2-32 Rev. 2

APR1400 DCD TIER 2 Table 12.2-9 Spent Fuel Gamma Source Time after Shutdown 50 hr 100 hr 200 hr 500 hr 1,000 hr Mean Energy (MeV) MeV/W-s 1.00E-02 2.04E+08 1.31E+08 6.94E+07 3.50E+07 2.60E+07 3.00E-02 1.62E+08 1.25E+08 8.86E+07 5.23E+07 3.60E+07 5.50E-02 1.37E+08 1.02E+08 7.08E+07 4.36E+07 3.16E+07 8.50E-02 2.69E+08 1.93E+08 1.15E+08 5.05E+07 3.07E+07 1.20E-01 1.11E+09 6.38E+08 2.37E+08 5.87E+07 4.38E+07 1.70E-01 3.10E+08 2.51E+08 1.91E+08 1.30E+08 8.84E+07 3.00E-01 2.21E+09 1.39E+09 6.79E+08 2.22E+08 1.04E+08 6.50E-01 8.52E+09 7.02E+09 5.77E+09 4.40E+09 3.46E+09 1.13E+00 1.28E+09 9.12E+08 5.90E+08 2.91E+08 1.44E+08 1.58E+00 3.16E+09 2.81E+09 2.22E+09 1.12E+09 3.84E+08 2.00E+00 3.20E+08 2.72E+08 2.12E+08 1.23E+08 6.18E+07 2.40E+00 2.38E+08 2.13E+08 1.72E+08 8.80E+07 3.04E+07 2.80E+00 4.13E+06 3.62E+06 2.95E+06 1.69E+06 8.20E+05 3.25E+00 1.87E+06 1.71E+06 1.40E+06 7.50E+05 2.94E+05 3.75E+00 6.66E+01 5.27E+01 5.24E+01 5.12E+01 4.94E+01 4.25E+00 9.58E+00 6.75E+00 6.72E+00 6.63E+00 6.49E+00 4.75E+00 2.77E+01 4.38E+00 4.36E+00 4.30E+00 4.20E+00 5.50E+00 4.77E+00 4.59E+00 4.57E+00 4.50E+00 4.40E+00 12.2-33 Rev. 2

APR1400 DCD TIER 2 Table 12.2-10 (1 of 2)

CVCS Heat Exchanger Inventories, Maximum Values (Bq)

Nuclide Letdown Regenerative CCP Miniflow H-3 3.5E+10 2.6E+10 1.9E+10 N-16 1.4E+11 7.5E+10 0.0E+00 Kr-85m 2.8E+09 2.0E+09 1.5E+09 Kr-85 1.2E+10 8.9E+09 6.5E+09 Kr-87 2.2E+09 1.5E+09 1.1E+09 Kr-88 6.1E+09 4.3E+09 3.2E+09 Xe-131m 1.2E+10 8.9E+09 6.5E+09 Xe-133m 7.3E+08 5.4E+08 3.9E+08 Xe-133 7.8E+11 5.8E+11 4.2E+11 Xe-135m 1.6E+09 9.3E+08 6.5E+08 Xe-135 1.6E+10 1.2E+10 8.5E+09 Xe-137 3.7E+08 1.8E+08 1.2E+08 Xe-138 1.4E+09 8.0E+08 5.6E+08 Br-84 5.3E+07 5.2E+06 2.4E+05 Rb-88 6.2E+09 2.1E+09 1.3E+09 Sr-89 8.8E+06 9.3E+05 9.5E+04 Sr-90 5.9E+05 6.2E+04 6.4E+03 Sr-91 1.3E+07 1.4E+06 1.4E+05 Y-91m 7.6E+06 5.0E+06 3.6E+06 Y-91 1.3E+06 9.6E+05 7.0E+05 Y-93 3.1E+05 2.3E+05 1.7E+05 Zr-95 4.2E+06 4.4E+05 4.5E+04 Nb-95 1.4E+06 1.5E+05 1.5E+04 Tc-99m 4.4E+08 4.6E+07 4.7E+06 Mo-99 7.6E+08 8.0E+07 8.2E+06 Ru-103 4.7E+05 5.0E+04 5.1E+03 Ru-106 2.0E+05 2.1E+04 2.2E+03 Ag-110m 1.4E+07 1.5E+06 1.5E+05 12.2-34 Rev. 2

APR1400 DCD TIER 2 Table 12.2-10 (2 of 2)

Nuclide Letdown Regenerative CCP Miniflow Te-129m 1.6E+07 1.6E+06 8.6E+04 Te-129 1.7E+07 1.7E+06 8.3E+04 I-131 6.7E+09 6.6E+08 3.6E+07 Te-131m 7.6E+07 7.5E+06 4.1E+05 Te-131 3.0E+07 2.9E+06 1.3E+05 Te-132 5.3E+08 5.2E+07 2.9E+06 I-132 1.8E+09 1.8E+08 9.2E+06 I-133 9.5E+09 9.4E+08 5.1E+07 I-134 1.1E+09 1.1E+08 5.3E+06 Cs-134 9.5E+08 4.0E+08 2.6E+08 I-135 5.4E+09 5.3E+08 2.9E+07 Cs-136 1.3E+08 5.4E+07 3.5E+07 Cs-137 1.1E+09 4.6E+08 3.0E+08 Ba-140 1.1E+07 1.2E+06 1.2E+05 La-140 3.7E+06 3.9E+05 4.0E+04 Ce-141 4.1E+05 4.3E+04 4.4E+03 Ce-143 1.1E+06 1.2E+05 1.2E+04 Ce-144 1.2E+06 1.3E+05 1.3E+04 Na-24 4.9E+08 5.1E+07 5.2E+06 Cr-51 1.5E+08 1.6E+07 1.6E+06 Mn-54 1.7E+07 1.8E+06 1.8E+05 Fe-55 1.3E+07 1.4E+06 1.4E+05 Fe-59 3.2E+06 3.4E+05 3.5E+04 Co-58 4.9E+07 5.2E+06 5.3E+05 Co-60 5.7E+06 6.0E+05 6.1E+04 Zn-65 5.5E+06 5.8E+05 5.9E+04 (1)

Ba-137m 1.1E+09 4.6E+08 3.0E+08 W-187 2.6E+07 2.8E+06 2.8E+05 Np-239 2.3E+07 2.5E+06 2.5E+05 (1) This nuclide is a daughter nuclide in secular equilibrium and the activity is that of the parent nuclide (Cs-137).

12.2-35 Rev. 2

APR1400 DCD TIER 2 Table 12.2-11 (1 of 2)

CVCS Ion Exchanger Inventories, Maximum Values (Bq)

Nuclide Purification Deborating Pre-Holdup Boric Acid Condensate H-3 1.5E+11 1.5E+11 1.4E+11 1.3E+11 N-16 5.4E+08 0.0E+00 0.0E+00 0.0E+00 Kr-85m 1.2E+10 1.2E+10 1.1E+10 1.4E+02 Kr-85 5.0E+10 5.0E+10 4.9E+10 3.1E+04 Kr-87 9.2E+09 9.2E+09 8.6E+09 3.2E+01 Kr-88 2.5E+10 2.5E+10 2.4E+10 2.0E+02 Xe-131m 5.0E+10 5.0E+10 4.8E+10 1.6E+04 Xe-133m 3.0E+09 3.0E+09 2.9E+09 4.1E+02 Xe-133 3.3E+12 3.3E+12 3.1E+12 6.0E+05 Xe-135m 6.7E+09 6.7E+09 6.2E+09 5.0E+00 Xe-135 6.7E+10 6.7E+10 6.2E+10 1.7E+03 Xe-137 1.5E+09 1.5E+09 1.4E+09 3.0E-01 Xe-138 5.8E+09 5.8E+09 5.4E+09 3.9E+00 Br-84 2.7E+09 2.5E+07 3.1E+06 5.5E-01 Rb-88 1.0E+11 1.3E+10 1.6E+09 1.0E+02 Sr-89 1.0E+12 7.3E+05 2.3E+09 3.1E+02 Sr-90 4.4E+11 4.9E+04 1.0E+09 2.5E+01 Sr-91 1.2E+10 1.1E+06 1.7E+07 1.1E+01 Y-91m 3.2E+07 3.2E+07 3.0E+07 8.1E+01 Y-91 5.4E+06 5.4E+06 5.3E+06 4.2E+03 Y-93 1.3E+06 1.3E+06 1.2E+06 3.9E+01 Zr-95 6.2E+11 3.5E+05 1.4E+09 1.5E+02 Nb-95 1.1E+11 1.2E+05 2.5E+08 4.5E+01 Tc-99m 2.5E+11 3.7E+07 3.4E+08 2.4E+02 Mo-99 4.8E+12 6.3E+07 8.2E+09 4.8E+03 Ru-103 4.3E+10 3.9E+04 9.5E+07 1.6E+01 Ru-106 1.0E+11 1.7E+04 2.3E+08 8.2E+00 Ag-110m 5.9E+12 1.2E+06 1.4E+10 5.6E+02 12.2-36 Rev. 2

APR1400 DCD TIER 2 Table 12.2-11 (2 of 2)

Nuclide Purification Deborating Pre-Holdup Boric Acid Condensate Te-129m 1.3E+12 4.1E+09 2.6E+09 1.9E+05 Te-129 1.9E+09 1.7E+07 2.2E+06 1.2E+05 (2)

I-131 1.2E+14 9.6E+11 2.2E+11 9.3E+06 Te-131m 2.1E+11 1.9E+09 2.9E+08 2.5E+03 Te-131 1.2E+09 1.1E+07 1.4E+06 5.6E+02 (2)

Te-132 4.0E+12 3.6E+10 6.3E+09 1.3E+05 I-132 4.0E+11 3.6E+09 4.6E+08 1.3E+05 (2)

I-133 1.9E+13 1.7E+11 2.5E+10 1.6E+05 I-134 9.2E+10 8.4E+08 1.1E+08 3.1E+01 Cs-134 3.5E+14 2.0E+09 6.1E+12 1.6E+04 I-135 3.5E+12 3.2E+10 4.2E+09 9.2E+03 Cs-136 2.3E+12 2.7E+08 3.9E+10 1.3E+03 Cs-137 5.0E+14 2.3E+09 8.6E+12 1.9E+04 Ba-140 3.2E+11 9.2E+05 6.6E+08 2.4E+02 La-140 1.4E+10 3.1E+05 2.3E+07 1.4E+01 Ce-141 3.1E+10 3.4E+04 6.7E+07 1.3E+01 Ce-143 3.5E+09 9.2E+04 5.4E+06 3.4E+00 Ce-144 5.4E+11 1.0E+05 1.2E+09 4.9E+01 Na-24 7.0E+11 4.1E+07 1.0E+09 6.7E+02 Cr-51 9.4E+12 7.2E+10 2.1E+10 4.5E+03 Mn-54 8.0E+12 1.1E+10 1.9E+10 7.0E+02 Fe-55 8.3E+12 8.0E+09 1.9E+10 5.3E+02 Fe-59 3.3E+11 1.7E+09 7.4E+08 1.1E+02 Co-58 8.0E+12 2.8E+10 1.8E+10 1.8E+03 Co-60 4.0E+12 3.6E+09 9.2E+09 2.4E+02 Zn-65 2.3E+12 4.5E+05 5.2E+09 2.2E+02 Ba-137m(1) 5.0E+14 2.3E+09 8.6E+12 1.9E+04 W-187 6.0E+10 2.2E+06 8.9E+07 5.8E+01 Np-239 1.3E+11 1.9E+06 2.1E+08 1.2E+02 (1) This nuclide is a daughter nuclide in secular equilibrium and the activity is that of the parent nuclide (Cs-137).

(2) The inventories include the contributions from the decay of parents to cover the potential non-conservatism of not including the buildup of daughters due to the decay of Telluriums.

12.2-37 Rev. 2

APR1400 DCD TIER 2 Table 12.2-12 (1 of 2)

CVCS Filter Inventories, Maximum Values (Bq)

Seal Reactor Boric Reactor Nuclide Injection Drain Acid Purification Makeup Water H-3 9.8E+08 1.7E+09 1.6E+09 2.0E+09 1.9E+09 N-16 0.0E+00 0.0E+00 0.0E+00 1.8E+08 0.0E+00 Kr-85m 7.7E+07 6.9E+07 0.0E+00 1.6E+08 7.1E-03 Kr-85 3.4E+08 5.7E+08 2.7E+04 6.7E+08 2.0E+02 Kr-87 5.6E+07 5.3E+07 0.0E+00 1.2E+08 4.6E-04 Kr-88 1.6E+08 1.5E+08 0.0E+00 3.4E+08 6.2E-03 Xe-131m 3.4E+08 4.9E+08 3.3E+02 6.7E+08 4.8E+01 Xe-133m 2.0E+07 2.3E+07 2.2E-02 4.1E+07 2.5E-01 Xe-133 2.2E+10 2.9E+10 1.5E+03 4.4E+10 8.2E+02 Xe-135m 3.4E+07 3.8E+07 0.0E+00 9.0E+07 1.5E-05 Xe-135 4.4E+08 4.1E+08 0.0E+00 9.0E+08 1.7E-01 Xe-137 6.3E+06 8.8E+06 0.0E+00 2.1E+07 2.4E-07 Xe-138 2.9E+07 3.3E+07 0.0E+00 7.8E+07 1.0E-05 Br-84 1.2E+04 1.3E+06 0.0E+00 3.0E+06 1.3E-06 Rb-88 6.6E+07 1.5E+08 0.0E+00 3.5E+08 3.4E-04 Sr-89 4.9E+03 4.0E+05 5.4E+04 4.9E+05 2.4E+00 Sr-90 3.3E+02 2.8E+04 3.2E+04 3.3E+04 3.7E-01 Sr-91 7.2E+03 3.3E+05 1.6E-12 7.3E+05 1.2E-03 Y-91m 1.9E+05 1.8E+05 0.0E+00 4.3E+05 7.6E-04 Y-91 3.6E+04 6.0E+04 8.6E+05 7.3E+04 3.5E+01 Y-93 8.6E+03 8.0E+03 1.5E-11 1.7E+04 4.4E-03 Zr-95 2.3E+03 1.9E+05 3.5E+04 2.3E+05 1.3E+00 Nb-95 7.8E+02 6.3E+04 5.0E+03 7.8E+04 2.9E-01 Tc-99m 2.4E+05 1.1E+07 0.0E+00 2.5E+07 1.6E-02 Mo-99 4.2E+05 2.5E+07 7.8E+02 4.3E+07 3.5E+00 Ru-103 2.6E+02 2.1E+04 2.0E+03 2.6E+04 1.1E-01 Ru-106 1.1E+02 9.5E+03 6.9E+03 1.1E+04 1.1E-01 Ag-110m 7.8E+03 6.6E+05 4.0E+05 7.8E+05 7.1E+00 12.2-38 Rev. 2

APR1400 DCD TIER 2 Table 12.2-12 (2 of 2)

Seal Reactor Boric Reactor Nuclide Injection Drain Acid Purification Makeup Water Te-129m 4.5E+03 7.2E+05 5.1E+04 9.0E+05 3.1E-01 Te-129 4.3E+03 4.1E+05 0.0E+00 9.5E+05 1.9E-06 I-131 1.9E+06 2.6E+08 8.4E+05 3.8E+08 2.1E+01 Te-131m 2.1E+04 2.2E+06 4.1E-02 4.3E+06 5.9E-03 Te-131 6.8E+03 7.2E+05 0.0E+00 1.7E+06 4.6E-07 Te-132 1.5E+05 1.8E+07 1.3E+03 3.0E+07 3.1E-01 I-132 4.8E+05 4.4E+07 0.0E+00 1.0E+08 8.3E-04 I-133 2.6E+06 2.6E+08 5.3E-02 5.3E+08 3.8E-01 I-134 2.7E+05 2.6E+07 0.0E+00 6.2E+07 7.3E-05 Cs-134 1.3E+07 4.5E+07 1.7E+07 5.3E+07 2.3E+02 I-135 1.5E+06 1.4E+08 0.0E+00 3.0E+08 2.2E-02 Cs-136 1.8E+06 5.4E+06 3.0E+04 7.3E+06 4.1E+00 Cs-137 1.5E+07 5.2E+07 2.5E+07 6.2E+07 2.8E+02 Ba-140 6.2E+03 4.6E+05 5.5E+03 6.2E+05 7.6E-01 La-140 2.1E+03 1.1E+05 7.8E-02 2.1E+05 6.3E-03 Ce-141 2.3E+02 1.8E+04 1.3E+03 2.3E+04 8.1E-02 Ce-143 6.1E+02 3.2E+04 3.1E-03 6.2E+04 1.2E-03 Ce-144 6.7E+02 5.7E+04 3.7E+04 6.7E+04 6.3E-01 Na-24 2.7E+05 1.3E+07 6.5E-06 2.7E+07 1.1E-01 Cr-51 5.7E+10 1.8E+10 8.5E+08 8.6E+12 5.3E+05 Mn-54 4.9E+10 1.6E+10 9.5E+09 7.4E+12 1.4E+06 Fe-55 5.0E+10 1.7E+10 1.4E+10 7.6E+12 1.6E+06 Fe-59 2.0E+09 6.4E+08 6.3E+07 3.0E+11 2.8E+04 Co-58 4.8E+10 1.6E+10 2.7E+09 7.3E+12 8.8E+05 Co-60 2.4E+10 8.0E+09 7.4E+09 3.6E+12 8.0E+05 Zn-65 3.1E+03 2.6E+05 1.5E+05 3.1E+05 2.8E+00 (1)

Ba-137m 1.5E+07 5.2E+07 2.5E+07 6.2E+07 2.8E+02 W-187 1.5E+04 7.4E+05 1.3E-03 1.5E+06 1.6E-02 Np-239 1.3E+04 7.5E+05 7.8E+00 1.3E+06 7.8E-02 (1) This nuclide is a daughter nuclide in secular equilibrium and the activity is that of the parent nuclide (Cs-137).

12.2-39 Rev. 2

APR1400 DCD TIER 2 Table 12.2-13 (1 of 5)

CVCS Tank Inventories, Maximum Values (Bq)

Reactor Drain Equipment Drain Volume Control Nuclide Liquid Vapor Liquid Vapor Liquid Vapor H-3 1.6E+12 7.9E+08 1.7E+11 3.7E+06 2.0E+12 1.0E+09 N-16 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 Kr-85m 6.4E+10 7.9E+10 1.1E+10 1.2E+10 1.6E+11 3.6E+12 Kr-85 5.3E+11 8.5E+12 4.7E+10 3.5E+11 7.0E+11 1.6E+13 Kr-87 4.9E+10 1.8E+10 8.5E+09 2.8E+09 1.2E+11 2.5E+12 Kr-88 1.4E+11 1.1E+11 2.4E+10 1.7E+10 3.4E+11 7.7E+12 Xe-131m 4.6E+11 3.3E+12 4.7E+10 1.3E+10 7.0E+11 8.8E+12 Xe-133m 2.2E+10 9.5E+10 3.0E+09 1.8E+10 4.2E+10 5.4E+11 Xe-133 2.7E+13 1.6E+14 3.0E+12 3.8E+11 4.5E+13 5.7E+14 Xe-135m 3.5E+10 1.5E+09 6.2E+09 2.3E+08 7.0E+10 6.0E+11 Xe-135 3.8E+11 4.9E+11 6.2E+10 7.8E+10 9.2E+11 1.1E+13 Xe-137 8.1E+09 9.3E+07 1.4E+09 1.4E+07 1.3E+10 5.8E+10 Xe-138 3.1E+10 1.2E+09 5.4E+09 1.8E+08 6.0E+10 4.9E+11 Br-84 1.2E+09 5.9E+03 2.0E+08 1.2E+01 2.6E+07 1.0E+04 Rb-88 1.4E+11 3.8E+05 2.4E+10 7.7E+02 1.4E+11 4.8E+07 Sr-89 3.7E+08 1.8E+05 4.2E+07 7.8E+02 1.0E+07 5.1E+03 Sr-90 2.6E+07 1.3E+04 2.9E+06 6.3E+01 6.9E+05 3.4E+02 Sr-91 3.1E+08 2.4E+04 5.1E+07 5.1E+01 1.5E+07 7.4E+03 Y-91m 1.7E+08 1.3E+03 2.9E+07 2.7E+00 3.9E+08 1.7E+05 Y-91 5.5E+07 2.7E+04 6.2E+06 1.2E+02 7.6E+07 3.8E+04 Y-93 7.4E+06 6.0E+02 1.2E+06 1.3E+00 1.8E+07 8.8E+03 Zr-95 1.8E+08 8.7E+04 2.0E+07 3.8E+02 4.8E+06 2.4E+03 Nb-95 5.8E+07 2.8E+04 6.5E+06 1.1E+02 1.6E+06 8.1E+02 Tc-99m 1.0E+10 5.3E+05 1.7E+09 1.1E+03 5.0E+08 2.5E+05 Mo-99 2.3E+10 6.5E+06 3.1E+09 1.7E+04 8.8E+08 4.4E+05 Ru-103 2.0E+07 9.5E+03 2.2E+06 4.0E+01 5.5E+05 2.7E+02 Ru-106 8.8E+06 4.4E+03 9.8E+05 2.1E+01 2.3E+05 1.2E+02 Ag-110m 6.1E+08 3.1E+05 6.8E+07 1.4E+03 1.6E+07 8.1E+03 12.2-40 Rev. 2

APR1400 DCD TIER 2 Table 12.2-13 (2 of 5)

Reactor Drain Equipment Drain Volume Control Nuclide Liquid Vapor Liquid Vapor Liquid Vapor Te-129m 6.7E+08 3.2E+05 7.5E+07 1.3E+03 9.3E+06 4.6E+03 Te-129 3.8E+08 4.0E+03 6.6E+07 8.1E+00 9.0E+06 4.0E+03 I-131 2.4E+11 9.7E+07 2.9E+10 3.1E+05 3.9E+09 1.9E+06 Te-131m 2.0E+09 3.6E+05 3.0E+08 8.3E+02 4.4E+07 2.2E+04 Te-131 6.6E+08 2.6E+03 1.2E+08 5.3E+00 1.4E+07 5.4E+03 Te-132 1.7E+10 5.0E+06 2.2E+09 1.3E+04 3.1E+08 1.5E+05 I-132 4.0E+10 8.5E+05 7.0E+09 1.7E+03 9.9E+08 4.7E+05 I-133 2.4E+11 3.5E+07 3.7E+10 7.7E+04 5.5E+09 2.7E+06 I-134 2.4E+10 2.0E+05 4.2E+09 4.0E+02 5.7E+08 2.5E+05 Cs-134 4.2E+10 2.1E+07 4.7E+09 1.0E+05 2.8E+10 1.4E+07 I-135 1.3E+11 7.2E+06 2.1E+10 1.5E+04 3.1E+09 1.5E+06 Cs-136 5.0E+09 2.2E+06 5.8E+08 7.6E+03 3.8E+09 1.9E+06 Cs-137 4.9E+10 2.5E+07 5.4E+09 1.2E+05 3.2E+10 1.6E+07 Ba-140 4.2E+08 1.8E+05 4.9E+07 6.4E+02 1.3E+07 6.4E+03 La-140 1.0E+08 2.3E+04 1.5E+07 5.4E+01 4.3E+06 2.1E+03 Ce-141 1.7E+07 8.1E+03 1.9E+06 3.3E+01 4.8E+05 2.4E+02 Ce-143 3.0E+07 5.8E+03 4.4E+06 1.4E+01 1.3E+06 6.3E+02 Ce-144 5.3E+07 2.6E+04 5.9E+06 1.2E+02 1.4E+06 7.0E+02 Na-24 1.2E+10 1.4E+06 1.9E+09 2.9E+03 5.6E+08 2.8E+05 Cr-51 6.1E+09 2.9E+06 6.9E+08 1.1E+04 1.7E+08 8.6E+04 Mn-54 7.5E+08 3.8E+05 8.4E+07 1.8E+03 2.0E+07 1.0E+04 Fe-55 5.7E+08 2.9E+05 6.3E+07 1.4E+03 1.5E+07 7.5E+03 Fe-59 1.4E+08 6.6E+04 1.5E+07 2.8E+02 3.7E+06 1.9E+03 Co-58 2.1E+09 1.0E+06 2.4E+08 4.6E+03 5.7E+07 2.9E+04 Co-60 2.5E+08 1.3E+05 2.8E+07 6.0E+02 6.6E+06 3.3E+03 Zn-65 2.4E+08 1.2E+05 2.7E+07 5.6E+02 6.3E+06 3.2E+03 (1)

Ba-137m 4.9E+10 2.5E+07 5.4E+09 1.2E+05 3.2E+10 1.6E+07 W-187 6.8E+08 1.1E+05 1.0E+08 2.4E+02 3.0E+07 1.5E+04 Np-239 6.9E+08 1.8E+05 9.4E+07 4.5E+02 2.7E+07 1.3E+04 (1) This nuclide is a daughter nuclide in secular equilibrium and the activity is that of the parent nuclide (Cs-137).

12.2-41 Rev. 2

APR1400 DCD TIER 2 Table 12.2-13 (3 of 5)

Holdup Reactor Makeup Water Boric Acid Storage Nuclide Liquid Vapor Liquid Vapor Liquid Vapor H-3 2.3E+15 1.0E+09 1.8E+14 1.1E+10 9.7E+13 3.3E+09 N-16 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 Kr-85m 3.3E+08 8.1E+09 6.6E+02 4.2E+03 0.0E+00 0.0E+00 Kr-85 1.4E+10 1.8E+11 1.9E+07 1.1E+08 1.6E+09 1.3E+09 Kr-87 7.9E+07 1.9E+09 4.3E+01 2.7E+02 0.0E+00 0.0E+00 Kr-88 5.2E+08 1.1E+10 5.8E+02 3.7E+03 0.0E+00 0.0E+00 Xe-131m 8.6E+09 6.6E+09 4.5E+06 1.2E+07 2.0E+07 1.1E+08 Xe-133m 2.9E+08 8.7E+09 2.4E+04 6.4E+04 1.3E+03 7.6E+03 Xe-133 3.9E+11 1.9E+11 7.7E+07 2.1E+08 8.7E+07 5.0E+08 Xe-135m 4.7E+06 1.3E+08 1.4E+00 3.9E+00 0.0E+00 0.0E+00 Xe-135 2.6E+09 4.1E+10 1.6E+04 4.5E+04 2.5E-09 1.4E-08 Xe-137 1.6E+05 7.7E+06 2.2E-02 6.2E-02 0.0E+00 0.0E+00 Xe-138 3.4E+06 9.9E+07 9.7E-01 2.7E+00 0.0E+00 0.0E+00 Br-84 1.1E+06 3.5E+01 1.2E-01 6.1E-06 0.0E+00 0.0E+00 Rb-88 1.0E+08 1.6E+03 3.2E+01 1.6E-03 0.0E+00 0.0E+00 Sr-89 1.8E+08 2.5E+03 2.3E+05 1.1E+01 3.2E+09 1.5E+05 Sr-90 2.4E+07 2.0E+02 3.5E+04 2.1E+00 1.9E+09 6.5E+04 Sr-91 1.3E+07 1.7E+02 1.2E+02 5.8E-03 9.6E-08 0.0E+00 Y-91m 1.5E+08 1.3E+03 7.2E+01 3.6E-03 0.0E+00 0.0E+00 Y-91 9.6E+09 3.4E+04 3.3E+06 1.7E+02 5.1E+10 2.3E+06 Y-93 7.6E+07 5.9E+02 4.1E+02 2.0E-02 9.2E-07 0.0E+00 Zr-95 8.9E+07 1.2E+03 1.2E+05 6.4E+00 2.1E+09 9.3E+04 Nb-95 2.6E+07 3.8E+02 2.8E+04 1.3E+00 3.0E+08 1.4E+04 Tc-99m 2.3E+08 3.7E+03 1.5E+03 7.5E-02 0.0E+00 0.0E+00 Mo-99 5.0E+09 6.1E+04 3.3E+05 1.6E+01 4.6E+07 2.3E+03 Ru-103 8.9E+06 1.3E+02 1.0E+04 5.0E-01 1.2E+08 5.6E+03 Ru-106 6.6E+06 6.5E+01 1.0E+04 5.9E-01 4.1E+08 1.5E+04 Ag-110m 4.3E+08 4.5E+03 6.6E+05 3.8E+01 2.4E+10 9.0E+05 12.2-42 Rev. 2

APR1400 DCD TIER 2 Table 12.2-13 (4 of 5)

Holdup Reactor Makeup Water Boric Acid Storage Nuclide Liquid Vapor Liquid Vapor Liquid Vapor Te-129m 2.7E+08 4.0E+03 2.9E+04 1.4E+00 3.0E+09 1.4E+05 Te-129 9.6E+05 2.4E+01 1.8E-01 9.0E-06 0.0E+00 0.0E+00 I-131 7.0E+10 1.0E+06 2.0E+06 9.3E+01 5.0E+10 2.5E+06 Te-131m 2.5E+08 2.6E+03 5.5E+02 2.7E-02 2.4E+03 1.2E-01 Te-131 4.4E+05 1.5E+01 4.3E-02 2.1E-06 0.0E+00 0.0E+00 Te-132 3.6E+09 4.4E+04 2.9E+04 1.4E+00 7.7E+07 3.8E+03 I-132 2.3E+08 5.1E+03 7.8E+01 3.9E-03 0.0E+00 0.0E+00 I-133 2.2E+10 2.4E+05 3.6E+04 1.8E+00 3.2E+03 1.6E-01 I-134 4.4E+07 1.2E+03 6.9E+00 3.4E-04 0.0E+00 0.0E+00 Cs-134 8.8E+09 1.3E+05 2.1E+07 1.3E+03 1.0E+12 3.6E+07 I-135 2.3E+09 4.5E+04 2.0E+03 1.0E-01 0.0E+00 0.0E+00 Cs-136 7.2E+08 1.2E+04 3.8E+05 1.8E+01 1.8E+09 8.8E+04 Cs-137 1.1E+10 1.5E+05 2.6E+07 1.6E+03 1.5E+12 5.0E+07 Ba-140 1.5E+08 2.3E+03 7.1E+04 3.3E+00 3.3E+08 1.6E+04 La-140 1.8E+07 1.9E+02 5.9E+02 2.9E-02 4.6E+03 2.3E-01 Ce-141 7.3E+06 1.1E+02 7.6E+03 3.6E-01 7.8E+07 3.7E+03 Ce-143 4.5E+06 4.8E+01 1.2E+02 5.8E-03 1.9E+02 9.3E-03 Ce-144 3.8E+07 3.9E+02 5.9E+04 3.4E+00 2.2E+09 8.2E+04 Na-24 8.8E+08 1.0E+04 1.1E+04 5.2E-01 3.9E-01 1.9E-05 Cr-51 2.5E+09 3.8E+04 2.4E+06 1.1E+02 2.2E+10 1.0E+06 Mn-54 5.5E+08 5.6E+03 8.5E+05 4.9E+01 3.3E+10 1.2E+06 Fe-55 4.9E+08 4.2E+03 7.2E+05 4.3E+01 3.5E+10 1.2E+06 Fe-59 6.3E+07 9.1E+02 7.7E+04 3.8E+00 9.9E+08 4.5E+04 Co-58 1.1E+09 1.5E+04 1.5E+06 8.0E+01 2.7E+10 1.2E+06 Co-60 2.3E+08 1.9E+03 3.3E+05 2.0E+01 1.7E+10 5.9E+05 Zn-65 1.7E+08 1.8E+03 2.6E+05 1.5E+01 9.1E+09 3.5E+05 (1)

Ba-137m 1.1E+10 1.5E+05 2.6E+07 1.6E+03 1.5E+12 5.0E+07 W-187 8.1E+07 8.4E+02 1.5E+03 7.2E-02 7.9E+01 4.0E-03 Np-239 1.4E+08 1.6E+03 7.3E+03 3.6E-01 4.6E+05 2.3E+01 (1) This nuclide is a daughter nuclide in secular equilibrium and the activity is that of the parent nuclide (Cs-137).

12.2-43 Rev. 2

APR1400 DCD TIER 2 Table 12.2-13 (5 of 5)

In-containment Refueling Water In-containment Refueling Water Storage Storage Nuclide Liquid Vapor Nuclide Liquid Vapor H-3 1.7E+14 2.7E+10 Te-129m 1.6E+09 8.8E+04 N-16 0.0E+00 0.0E+00 Te-129 3.1E-03 0.0E+00 Kr-85m 5.1E+08 0.0E+00 I-131 3.6E+11 3.5E+05 Kr-85 1.3E+12 4.0E+09 Te-131m 1.6E+09 2.5E-03 Kr-87 2.2E+01 0.0E+00 Te-131 0.0E+00 0.0E+00 Kr-88 2.3E+07 0.0E+00 Te-132 2.2E+10 2.1E+02 Xe-131m 1.2E+12 3.6E+06 I-132 2.1E+05 0.0E+00 Xe-133m 5.0E+10 4.7E+01 I-133 1.4E+11 2.3E-03 Xe-133 7.0E+13 7.1E+06 I-134 7.0E-05 0.0E+00 Xe-135m 0.0E+00 0.0E+00 Cs-134 1.6E+12 2.3E+08 Xe-135 8.8E+10 0.0E+00 I-135 3.7E+09 0.0E+00 Xe-137 0.0E+00 0.0E+00 Cs-136 1.4E+10 2.0E+04 Xe-138 0.0E+00 0.0E+00 Cs-137 2.7E+12 4.1E+08 Br-84 0.0E+00 0.0E+00 Ba-140 6.6E+08 3.6E+03 Rb-88 0.0E+00 0.0E+00 La-140 1.1E+08 6.6E-03 Sr-89 1.5E+09 1.4E+05 Ce-141 4.1E+07 2.2E+03 Sr-90 3.3E+09 5.4E+05 Ce-143 2.7E+07 2.2E-04 Sr-91 3.6E+07 0.0E+00 Ce-144 2.4E+09 3.7E+05 Y-91m 4.0E-07 0.0E+00 Na-24 4.1E+09 2.1E-07 Y-91 2.0E+10 2.6E+06 Cr-51 1.3E+10 5.2E+05 Y-93 3.8E+06 0.0E+00 Mn-54 3.7E+10 5.8E+06 Zr-95 1.1E+09 1.2E+05 Fe-55 5.3E+10 8.5E+06 Nb-95 1.5E+08 9.0E+03 Fe-59 4.7E+08 3.8E+04 Tc-99m 1.8E+08 0.0E+00 Co-58 1.4E+10 1.7E+06 Mo-99 3.0E+10 1.1E+02 Co-60 2.8E+10 4.5E+06 Ru-103 5.9E+07 4.1E+03 Zn-65 9.2E+09 1.4E+06 Ru-106 4.9E+08 7.7E+04 Ba-137m 2.7E+12 4.1E+08 Ag-110m 2.4E+10 3.7E+06 W-187 4.6E+08 6.8E-05 Np-239 8.5E+08 9.3E-01 12.2-44 Rev. 2

APR1400 DCD TIER 2 Table 12.2-14 (1 of 2)

Boric Acid Concentrator Package Inventories, Maximum Values (Bq)

Concentrate Concentrate Concentrate Flash Vapor Concentrate Transfer Nuclide Heater Cooler Tank Separator Pump Pump H-3 6.9E+10 5.3E+09 6.0E+11 1.3E+08 5.8E+09 5.8E+08 N-16 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 Kr-85m 7.7E+03 5.9E+02 7.3E+05 2.0E+05 6.5E+02 6.5E+01 Kr-85 1.7E+06 1.3E+05 1.6E+08 4.2E+07 1.4E+05 1.4E+04 Kr-87 1.7E+03 1.3E+02 1.7E+05 4.4E+04 1.5E+02 1.5E+01 Kr-88 1.1E+04 8.2E+02 1.0E+06 2.7E+05 8.9E+02 8.9E+01 Xe-131m 8.8E+05 6.8E+04 8.4E+07 2.2E+07 7.4E+04 7.4E+03 Xe-133m 2.2E+04 1.7E+03 2.1E+06 5.6E+05 1.9E+03 1.9E+02 Xe-133 3.2E+07 2.5E+06 3.1E+09 8.2E+08 2.7E+06 2.7E+05 Xe-135m 2.7E+02 2.1E+01 2.6E+04 6.8E+03 2.3E+01 2.3E+00 Xe-135 9.0E+04 6.9E+03 8.5E+06 2.3E+06 7.5E+03 7.5E+02 Xe-137 1.6E+01 1.3E+00 1.6E+03 4.1E+02 1.4E+00 1.4E-01 Xe-138 2.1E+02 1.6E+01 2.0E+04 5.3E+03 1.8E+01 1.8E+00 Br-84 1.2E+05 9.1E+03 1.0E+06 2.3E+02 9.9E+03 9.9E+02 Rb-88 5.6E+06 4.3E+05 4.8E+07 1.1E+04 4.7E+05 4.7E+04 Sr-89 1.7E+07 1.3E+06 1.4E+08 3.2E+04 1.4E+06 1.4E+05 Sr-90 1.3E+06 1.0E+05 1.1E+07 2.6E+03 1.1E+05 1.1E+04 Sr-91 6.1E+05 4.7E+04 5.3E+06 1.2E+03 5.1E+04 5.1E+03 Y-91m 4.4E+06 3.4E+05 3.8E+07 8.4E+03 3.7E+05 3.7E+04 Y-91 2.2E+08 1.7E+07 1.9E+09 4.3E+05 1.9E+07 1.9E+06 Y-93 2.1E+06 1.6E+05 1.8E+07 4.0E+03 1.8E+05 1.8E+04 Zr-95 8.2E+06 6.3E+05 7.0E+07 1.6E+04 6.8E+05 6.8E+04 Nb-95 2.4E+06 1.9E+05 2.1E+07 4.7E+03 2.0E+05 2.0E+04 Tc-99m 1.3E+07 9.8E+05 1.1E+08 2.4E+04 1.1E+06 1.1E+05 Mo-99 2.6E+08 2.0E+07 2.2E+09 5.0E+05 2.2E+07 2.2E+06 Ru-103 8.4E+05 6.5E+04 7.3E+06 1.6E+03 7.1E+04 7.1E+03 Ru-106 4.4E+05 3.4E+04 3.8E+06 8.5E+02 3.7E+04 3.7E+03 Ag-110m 3.0E+07 2.3E+06 2.6E+08 5.8E+04 2.5E+06 2.5E+05 12.2-45 Rev. 2

APR1400 DCD TIER 2 Table 12.2-14 (2 of 2)

Concentrate Concentrate Concentrate Flash Vapor Concentrate Transfer Nuclide Heater Cooler Tank Separator Pump Pump Te-129m 2.6E+07 2.0E+06 2.2E+08 5.0E+04 2.2E+06 2.2E+05 Te-129 8.1E+04 6.2E+03 7.0E+05 1.6E+02 6.8E+03 6.8E+02 I-131 5.5E+09 4.2E+08 4.7E+10 1.0E+07 4.6E+08 4.6E+07 Te-131m 9.9E+06 7.6E+05 8.6E+07 1.9E+04 8.3E+05 8.3E+04 Te-131 5.2E+04 4.0E+03 4.5E+05 1.0E+02 4.4E+03 4.4E+02 Te-132 1.9E+08 1.5E+07 1.7E+09 3.7E+05 1.6E+07 1.6E+06 I-132 1.7E+07 1.3E+06 1.5E+08 3.3E+04 1.5E+06 1.5E+05 I-133 8.8E+08 6.8E+07 7.6E+09 1.7E+06 7.4E+07 7.4E+06 I-134 4.0E+06 3.1E+05 3.5E+07 7.8E+03 3.4E+05 3.4E+04 Cs-134 8.7E+08 6.7E+07 7.5E+09 1.7E+06 7.3E+07 7.3E+06 I-135 1.6E+08 1.2E+07 1.3E+09 3.0E+05 1.3E+07 1.3E+06 Cs-136 6.9E+07 5.3E+06 6.0E+08 1.3E+05 5.8E+06 5.8E+05 Cs-137 1.0E+09 7.8E+07 8.8E+09 2.0E+06 8.5E+07 8.5E+06 Ba-140 1.3E+07 1.0E+06 1.1E+08 2.5E+04 1.1E+06 1.1E+05 La-140 7.5E+05 5.8E+04 6.5E+06 1.5E+03 6.3E+04 6.3E+03 Ce-141 7.0E+05 5.4E+04 6.1E+06 1.4E+03 5.9E+04 5.9E+03 Ce-143 1.8E+05 1.4E+04 1.6E+06 3.5E+02 1.5E+04 1.5E+03 Ce-144 2.6E+06 2.0E+05 2.3E+07 5.0E+03 2.2E+05 2.2E+04 Na-24 3.6E+07 2.8E+06 3.1E+08 6.9E+04 3.0E+06 3.0E+05 Cr-51 2.4E+08 1.9E+07 2.1E+09 4.7E+05 2.0E+07 2.0E+06 Mn-54 3.8E+07 2.9E+06 3.2E+08 7.2E+04 3.2E+06 3.2E+05 Fe-55 2.9E+07 2.2E+06 2.5E+08 5.5E+04 2.4E+06 2.4E+05 Fe-59 5.9E+06 4.6E+05 5.1E+07 1.1E+04 5.0E+05 5.0E+04 Co-58 9.7E+07 7.5E+06 8.4E+08 1.9E+05 8.2E+06 8.2E+05 Co-60 1.3E+07 9.8E+05 1.1E+08 2.5E+04 1.1E+06 1.1E+05 Zn-65 1.2E+07 9.1E+05 1.0E+08 2.3E+04 1.0E+06 1.0E+05 (1)

Ba-137m 1.0E+09 7.8E+07 8.8E+09 2.0E+06 8.5E+07 8.5E+06 W-187 3.1E+06 2.4E+05 2.7E+07 6.0E+03 2.6E+05 2.6E+04 Np-239 6.7E+06 5.2E+05 5.8E+07 1.3E+04 5.6E+05 5.6E+04 (1) This nuclide is a daughter nuclide in secular equilibrium and the activity is that of the parent nuclide (Cs-137).

12.2-46 Rev. 2

APR1400 DCD TIER 2 Table 12.2-15 Gamma Sources of Irradiated ICI, CEA, NSA, and SCA In-Core Control Element Neutron Source Surveillance Mean Energy Instrument (1) Assembly Assembly Capsule Assembly (MeV) (/sec/cm) (/sec/cm) (/sec/assembly) (/sec/cm) 3.00E-01 3.72E+10 1.15E+10 6.90E+13 1.19E+09 6.50E-01 3.40E+11 2.43E+11 2.36E+15 1.04E+09 1.13E+00 1.08E+11 1.02E+11 2.46E+14 1.48E+10 1.58E+00 1.31E+09 6.61E+08 7.46E+14 2.26E+06 2.00E+00 5.76E+05 5.46E+05 8.02E+13 1.05E+05 2.40E+00 1.90E+03 1.74E+03 5.12E+11 1.24E+03 2.80E+00 7.20E+07 8.79E+00 5.14E+10 8.96E+02 3.25E+00 1.37E+01 1.61E+00 2.93E+06 1.64E+02 3.75E+00 4.68E+04 1.42E-08 3.25E+05 2.02E-08 4.25E+00 6.14E+02 0.00E+00 1.00E+05 2.64E-10 4.75E+00 0.00E+00 0.00E+00 5.70E+04 0.00E+00 5.50E+00 0.00E+00 0.00E+00 5.16E+04 0.00E+00 (1) Includes detector and associated cable.

12.2-47 Rev. 2

APR1400 DCD TIER 2 Table 12.2-16 Shutdown Cooling System Specific Source Strength Energy (MeV) 0.25 0.50 0.75 1.00 1.38 2.00 3.00 4.00 Decay Time (hr) Maximum Values (/g-sec) 1 1.2E+06 4.9E+04 2.9E+04 1.2E+04 1.6E+04 1.6E+04 2.5E+03 5.4E+00 10 1.1E+06 3.1E+04 1.4E+04 3.6E+03 6.0E+03 2.3E+03 1.1E+03 4.2E-05 100 6.8E+05 6.1E+03 9.2E+03 5.3E+02 3.1E+02 8.6E-01 1.7E+01 0.0E+00 12.2-48 Rev. 2

APR1400 DCD TIER 2 Table 12.2-17 Fission and Corrosion Product Activities in the Spent Fuel Pool Nuclide Activity (Bq/g) Nuclide Activity (Bq/g)

H-3 4.0E+04 Te-129 0.0E+00 N-16 0.0E+00 I-131 1.6E+02 Kr-85m 0.0E+00 Te-131m 7.3E-01 Kr-85 0.0E+00 Te-131 0.0E+00 Kr-87 0.0E+00 Te-132 9.9E+00 Kr-88 0.0E+00 I-132 9.4E-05 Xe-131m 0.0E+00 I-133 6.1E+01 Xe-133m 0.0E+00 I-134 0.0E+00 Xe-133 0.0E+00 Cs-134 8.1E+01 Xe-135m 0.0E+00 I-135 1.7E+00 Xe-135 0.0E+00 Cs-136 6.2E+00 Xe-137 0.0E+00 Cs-137 1.1E+02 Xe-138 0.0E+00 Ba-140 2.8E-01 Br-84 0.0E+00 La-140 4.9E-02 Rb-88 0.0E+00 Ce-141 1.1E-02 Sr-89 2.6E-01 Ce-143 1.2E-02 Sr-90 8.5E-02 Ce-144 8.4E-02 Sr-91 1.6E-02 Na-24 1.8E+00 Y-91m 0.0E+00 Cr-51 4.0E+00 Y-91 6.5E-01 Mn-54 1.3E+00 Y-93 1.7E-03 Fe-55 1.5E+00 Zr-95 1.3E-01 Fe-59 9.4E-02 Nb-95 3.9E-02 Co-58 1.6E+00 Tc-99m 8.1E-02 Co-60 7.4E-01 Mo-99 1.3E+01 Zn-65 3.5E-01 Ru-103 1.3E-02 Ba-137m(1) 1.1E+02 Ru-106 1.6E-02 W-187 2.1E-01 Ag-110m 9.1E-01 Np-239 3.8E-01 Te-129m 4.4E-01 (1) This nuclide is a daughter nuclide in secular equilibrium and the activity is that of the parent nuclide (Cs-137).

12.2-49 Rev. 2

APR1400 DCD TIER 2 Table 12.2-17a SFP Demineralizer and Filter Source Terms Nuclide Demin. (Bq) Filter (Bq) Nuclide Demin. (Bq) Filter (Bq)

BR-84 0.00E+00 0.00E+00 RU-106 6.00E+07 0.00E+00 I-131 2.83E+11 0.00E+00 AG-110M 3.39E+09 0.00E+00 I-132 1.96E-23 0.00E+00 TE-129M 1.40E+09 0.00E+00 I-133 1.89E+08 0.00E+00 TE-129 0.00E+00 0.00E+00 I-134 0.00E+00 0.00E+00 TE-131M 1.89E+07 0.00E+00 I-135 1.33E+00 0.00E+00 TE-131 0.00E+00 0.00E+00 RB-88 0.00E+00 0.00E+00 TE-132 5.56E+09 0.00E+00 CS-134 2.78E+11 0.00E+00 BA-137M 3.81E+11 0.00E+00 CS-136 1.33E+10 0.00E+00 BA-140 6.56E+08 0.00E+00 CS-137 3.81E+11 0.00E+00 LA-140 4.51E+06 0.00E+00 NA-24 3.18E+05 0.00E+00 CE-141 3.48E+07 0.00E+00 SR-89 8.80E+08 0.00E+00 CE-143 4.92E+05 0.00E+00 SR-90 3.24E+08 0.00E+00 CE-144 3.14E+08 0.00E+00 SR-91 1.16E+01 0.00E+00 W-187 1.53E+06 0.00E+00 Y-91M 0.00E+00 0.00E+00 NP-239 1.02E+08 0.00E+00 Y-91 2.23E+09 0.00E+00 CR-51 1.22E+10 1.21E+10 Y-93 2.79E+00 0.00E+00 MN-54 4.86E+09 4.88E+09 ZR-95 4.51E+08 4.50E+08 FE-55 5.69E+09 5.71E+09 NB-95 1.25E+08 0.00E+00 FE-59 3.12E+08 3.11E+08 MO-99 5.31E+09 0.00E+00 CO-58 5.59E+09 5.59E+09 TC-99M 4.46E-03 0.00E+00 CO-60 2.82E+09 2.82E+09 RU-103 4.24E+07 0.00E+00 ZN-65 1.30E+09 1.30E+09 12.2-50 Rev. 2

APR1400 DCD TIER 2 Table 12.2-18 (1 of 3)

Steam Generator Blowdown and Condensate Polishing System Source Terms (0.25 % Fuel Defect)

SG Water SG Steam Blowdown Blowdown Blowdown Post- Condensate CPS Cation CPS Mixed Flash Isotope (Bq/g) (Bq/g) Mixed Bed (Bq) Pre-Filter (Bq) Filter (Bq) (Bq/g) Bed (Bq) Bed (Bq) Tank (Bq)

Kr-85m 0.00E+00 1.29E-01 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr-85 0.00E+00 5.49E-01 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr-87 0.00E+00 1.01E-01 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr-88 0.00E+00 2.80E-01 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe-131m 0.00E+00 5.49E-01 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe-133m 0.00E+00 3.34E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe-133 0.00E+00 3.58E+01 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe-135m 0.00E+00 7.32E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe-135 0.00E+00 7.32E-01 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe-137 0.00E+00 1.70E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe-138 0.00E+00 6.41E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Br-84 7.65E-02 7.65E-04 4.71E+06 0.00E+00 0.00E+00 2.43E-04 0.00E+00 1.28E+05 7.48E+05 I-131 2.99E+01 2.99E-01 6.73E+11 0.00E+00 0.00E+00 9.50E-02 0.00E+00 1.71E+10 2.92E+08 I-132 5.44E+00 5.44E-02 1.46E+09 0.00E+00 0.00E+00 1.73E-02 0.00E+00 3.97E+07 5.32E+07 I-133 4.05E+01 4.05E-01 9.90E+10 0.00E+00 0.00E+00 1.29E-01 0.00E+00 2.70E+09 3.96E+08 I-134 2.16E+00 2.16E-02 2.18E+08 0.00E+00 0.00E+00 6.87E-03 0.00E+00 5.94E+06 2.11E+07 I-135 2.08E+01 2.08E-01 1.63E+10 0.00E+00 0.00E+00 6.61E-02 0.00E+00 4.44E+08 2.03E+08 Rb-88 5.96E+00 2.98E-02 2.06E+08 0.00E+00 0.00E+00 1.36E-02 4.00E+06 4.00E+05 5.83E+07 12.2-51 Rev. 2

APR1400 DCD TIER 2 Table 12.2-18 (2 of 3)

SG Water SG Steam Blowdown Blowdown Blowdown Post- Condensate CPS Cation CPS Mixed Flash Isotope (Bq/g) (Bq/g) Mixed Bed (Bq) Pre-Filter (Bq) Filter (Bq) (Bq/g) Bed (Bq) Bed (Bq) Tank (Bq)

Cs-134 4.77E+00 2.39E-02 1.55E+12 0.00E+00 0.00E+00 1.08E-02 5.27E+08 5.63E+08 4.67E+07 Cs-136 6.50E-01 3.25E-03 2.36E+10 0.00E+00 0.00E+00 1.48E-03 6.65E+07 3.75E+07 6.35E+06 Cs-137 5.52E+00 2.76E-02 1.94E+12 0.00E+00 0.00E+00 1.25E-02 6.10E+08 6.61E+08 5.40E+07 Cr-51 6.78E-01 3.39E-03 5.20E+10 4.73E+10 4.73E+08 1.54E-03 7.23E+07 5.60E+07 9.87E+08 Mn-54 7.85E-02 3.93E-04 2.28E+10 2.07E+10 2.07E+08 7.14E-05 3.46E+06 3.63E+06 5.94E+10 Fe-55 5.88E-02 2.94E-04 1.95E+10 1.77E+10 1.77E+08 5.35E-05 2.60E+06 2.79E+06 6.80E+09 Fe-59 1.47E-02 7.36E-05 1.73E+09 1.57E+09 1.57E+07 1.34E-05 6.37E+05 5.57E+05 6.89E+10 Co-58 2.25E-01 1.13E-03 3.71E+10 3.37E+10 3.37E+08 2.05E-04 9.82E+06 9.28E+06 8.52E+09 Co-60 2.60E-02 1.30E-04 8.90E+09 8.09E+09 8.09E+07 2.36E-05 1.15E+06 1.24E+06 1.07E+09 Zr-95 1.91E-02 9.53E-05 2.96E+09 2.69E+09 2.69E+07 1.73E-05 8.30E+05 7.75E+05 8.03E+08 Zn-65 2.50E-02 1.25E-04 6.90E+09 6.27E+09 6.27E+07 2.27E-05 1.10E+06 1.15E+06 1.91E+08 N-16 7.38E-01 3.69E-03 7.09E+03 0.00E+00 0.00E+00 6.71E-04 5.52E+01 5.52E+00 7.22E+06 Na-24 2.08E+00 1.04E-02 3.64E+09 0.00E+00 0.00E+00 1.89E-03 2.72E+07 2.83E+06 2.04E+07 Sr-89 4.03E-02 2.02E-04 5.35E+09 0.00E+00 0.00E+00 3.67E-05 1.75E+06 1.58E+06 3.94E+05 Sr-90 2.70E-03 1.35E-05 9.48E+08 0.00E+00 0.00E+00 2.45E-06 1.19E+05 1.29E+05 2.64E+04 Sr-91 5.33E-02 2.67E-04 6.01E+07 0.00E+00 0.00E+00 4.85E-05 4.65E+05 4.68E+04 5.21E+05 Y-91m 1.47E-02 7.36E-05 1.43E+06 0.00E+00 0.00E+00 1.34E-05 1.11E+04 1.11E+03 1.44E+05 Y-91 5.95E-03 2.98E-05 8.64E+08 0.00E+00 0.00E+00 5.41E-06 2.59E+05 2.38E+05 5.82E+04 Y-93 1.28E-03 6.41E-06 1.52E+06 0.00E+00 0.00E+00 1.16E-06 1.18E+04 1.19E+03 1.25E+04 Nb-95 6.41E-03 3.20E-05 6.11E+08 0.00E+00 0.00E+00 5.82E-06 2.75E+05 2.28E+05 6.27E+04 12.2-52 Rev. 2

APR1400 DCD TIER 2 Table 12.2-18 (3 of 3)

SG Water SG Steam Blowdown Blowdown Blowdown Post- Condensate CPS Cation CPS Mixed Flash Isotope (Bq/g) (Bq/g) Mixed Bed (Bq) Pre-Filter (Bq) Filter (Bq) (Bq/g) Bed (Bq) Bed (Bq) Tank (Bq)

Mo-99 3.42E+00 1.71E-02 2.68E+10 0.00E+00 0.00E+00 3.11E-03 1.08E+08 2.09E+07 3.35E+07 Tc-99m 1.70E+00 8.48E-03 1.18E+09 0.00E+00 0.00E+00 1.54E-03 9.18E+06 9.18E+05 1.66E+07 Ru-103 2.15E-03 1.08E-05 2.28E+08 0.00E+00 0.00E+00 1.96E-06 9.28E+04 7.91E+04 2.10E+04 Ru-106 9.16E-04 4.58E-06 2.74E+08 0.00E+00 0.00E+00 8.33E-07 4.04E+04 4.26E+04 8.96E+03 Ag-110m 6.38E-02 3.19E-04 1.77E+10 0.00E+00 0.00E+00 5.80E-05 2.81E+06 2.93E+06 6.24E+05 Te-129m 7.32E-02 3.66E-04 6.78E+09 0.00E+00 0.00E+00 6.66E-05 3.14E+06 2.58E+06 7.16E+05 Te-129 3.91E-02 1.96E-04 5.24E+06 0.00E+00 0.00E+00 3.56E-05 4.08E+04 4.08E+03 3.83E+05 Te-131m 3.35E-01 1.68E-03 1.17E+09 0.00E+00 0.00E+00 3.05E-04 7.33E+06 9.12E+05 3.28E+06 Te-131 3.69E-02 1.85E-04 1.79E+06 0.00E+00 0.00E+00 3.35E-05 1.40E+04 1.40E+03 3.61E+05 Te-132 2.39E+00 1.20E-02 2.18E+10 0.00E+00 0.00E+00 2.18E-03 7.87E+07 1.69E+07 2.34E+07 Ba-137m 5.52E+00 2.76E-02 1.94E+12 0.00E+00 0.00E+00 1.25E-02 6.10E+08 6.61E+08 5.40E+07 Ba-140 5.02E-02 2.51E-04 1.79E+09 0.00E+00 0.00E+00 4.57E-05 2.05E+06 1.15E+06 4.91E+05 La-140 1.65E-02 8.25E-05 7.74E+07 0.00E+00 0.00E+00 1.50E-05 4.23E+05 6.03E+04 1.61E+05 Ce-141 1.88E-03 9.40E-06 1.70E+08 0.00E+00 0.00E+00 1.71E-06 8.07E+04 6.57E+04 1.84E+04 Ce-143 4.87E-03 2.44E-05 1.87E+07 0.00E+00 0.00E+00 4.43E-06 1.13E+05 1.46E+04 4.77E+04 Ce-144 5.50E-03 2.75E-05 1.57E+09 0.00E+00 0.00E+00 5.00E-06 2.42E+05 2.54E+05 5.38E+04 W-187 1.15E-01 5.73E-04 3.19E+08 0.00E+00 0.00E+00 1.04E-04 2.16E+06 2.48E+05 1.12E+06 Np-239 1.05E-01 5.23E-04 6.86E+08 0.00E+00 0.00E+00 9.51E-05 3.10E+06 5.34E+05 1.02E+06 12.2-53 Rev. 2

APR1400 DCD TIER 2 Table 12.2-19 (1 of 2)

Gaseous Radwaste System Source Terms (0.25 % Fuel Defect)

Buildup Activity on Charcoal Bed (1)

At Inlet Guard Bed 1st Delay Bed 2nd Delay Bed 3rd Delay Bed 4th Delay Bed At Outlet 3 3 3 Nuclide Ci//cm Bq/cm Ci Bq Ci Bq Ci Bq Ci Bq Ci Bq Ci//cm Bq/cm3 Kr-85m 8.70E+00 3.22E+05 2.24E+05 8.28E+09 9.92E+07 3.67E+12 3.85E+06 1.43E+11 1.50E+05 5.53E+09 5.81E+03 2.15E+08 1.98E-05 7.32E-01 Kr-85 3.73E+01 1.38E+06 9.60E+05 3.55E+10 1.44E+09 5.32E+13 1.44E+09 5.32E+13 1.44E+09 5.32E+13 1.44E+09 5.32E+13 3.73E+01 1.38E+06 Kr-87 6.71E+00 2.48E+05 1.72E+05 6.38E+09 2.26E+07 8.35E+11 2.41E+02 8.93E+06 2.58E-03 9.54E+01 2.76E-08 1.02E-03 8.75E-20 3.24E-15 Kr-88 1.89E+01 6.99E+05 4.86E+05 1.80E+10 1.41E+08 5.22E+12 8.37E+05 3.10E+10 4.97E+03 1.84E+08 2.95E+01 1.09E+06 2.34E-08 8.65E-04 Xe-131m 3.71E+01 1.37E+06 9.54E+05 3.53E+10 1.35E+10 4.99E+14 7.00E+09 2.59E+14 3.64E+09 1.35E+14 1.89E+09 6.98E+13 2.70E+00 9.98E+04 Xe-133m 2.23E+00 8.24E+04 5.73E+04 2.12E+09 3.01E+08 1.11E+13 8.55E+06 3.16E+11 2.43E+05 8.99E+09 6.90E+03 2.55E+08 1.45E-06 5.37E-02 Xe-133 2.40E+03 8.88E+07 6.17E+07 2.28E+12 6.19E+11 2.29E+16 1.40E+11 5.18E+15 3.16E+10 1.17E+15 7.16E+09 2.65E+14 6.27E+00 2.32E+05 Xe-135m 4.93E+00 1.82E+05 1.27E+05 4.69E+09 3.33E+06 1.23E+11 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe-135 4.94E+01 1.83E+06 1.27E+06 4.70E+10 1.19E+09 4.40E+13 1.36E+00 5.02E+04 1.55E-09 5.73E-05 1.77E-18 6.54E-14 8.39E-35 3.10E-30 Xe-137 1.13E+00 4.19E+04 2.91E+04 1.08E+09 1.91E+05 7.07E+09 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe-138 4.19E+00 1.55E+05 1.08E+05 3.99E+09 2.62E+06 9.69E+10 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Br-84 1.99E-06 7.37E-02 2.80E+00 1.03E+05 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 I-131 2.48E-04 9.16E+00 1.27E+05 4.68E+09 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 I-132 6.59E-05 2.44E+00 4.01E+02 1.48E+07 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 I-133 3.57E-04 1.32E+01 1.96E+04 7.27E+08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 I-134 4.17E-05 1.54E+00 9.67E+01 3.58E+06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 I-135 2.01E-04 7.44E+00 3.52E+03 1.30E+08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Rb-88 1.89E+01 6.99E+05 4.86E+05 1.80E+10 1.41E+08 5.22E+12 8.37E+05 3.10E+10 4.97E+03 1.84E+08 2.95E+01 1.09E+06 2.34E-08 8.65E-04 Cs-137 1.13E+00 4.19E+04 2.91E+04 1.08E+09 1.91E+05 7.07E+09 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ba-137m 1.13E+00 4.19E+04 2.91E+04 1.08E+09 1.91E+05 7.07E+09 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 (1) The continuous venting flowrates for the gas stripper, reactor drain tank, volume control tank and equipment drain tank are 0.32 scfm, 0.024 scfm, 0.004 scfm and 0.005 scfm, respectively.

12.2-54 Rev. 2

APR1400 DCD TIER 2 Table 12.2-19 (2 of 2)

Header Drain Tank Waste Gas Dryer Vessel for HEPA Filter Nuclide Ci Bq Ci Bq Ci Bq Kr-85m 4.93E+06 1.82E+11 1.68E+05 6.20E+09 1.65E+00 6.10E+04 Kr-85 2.11E+07 7.82E+11 7.19E+05 2.66E+10 3.11E+06 1.15E+11 Kr-87 3.80E+06 1.41E+11 1.29E+05 4.78E+09 7.30E-15 2.70E-10 Kr-88 1.07E+07 3.96E+11 3.64E+05 1.35E+10 1.95E-03 7.21E+01 Xe-131m 2.10E+07 7.77E+11 7.13E+05 2.64E+10 2.25E+05 8.32E+09 Xe-133m 1.26E+06 4.67E+10 4.29E+04 1.59E+09 1.21E-01 4.48E+03 Xe-133 1.36E+09 5.03E+13 4.62E+07 1.71E+12 5.23E+05 1.93E+10 Xe-135m 2.79E+06 1.03E+11 9.48E+04 3.51E+09 6.99E-30 2.59E-25 Xe-135 2.80E+07 1.04E+12 9.53E+05 3.53E+10 0.00E+00 0.00E+00 Xe-137 6.41E+05 2.37E+10 2.18E+04 8.07E+08 0.00E+00 0.00E+00 Xe-138 2.37E+06 8.78E+10 8.07E+04 2.99E+09 0.00E+00 0.00E+00 Br-84 1.13E+00 4.17E+04 3.84E-02 1.42E+03 0.00E+00 0.00E+00 I-131 1.40E+02 5.19E+06 4.77E+00 1.76E+05 0.00E+00 0.00E+00 I-132 3.73E+01 1.38E+06 1.27E+00 4.70E+04 0.00E+00 0.00E+00 I-133 2.02E+02 7.48E+06 6.87E+00 2.54E+05 0.00E+00 0.00E+00 I-134 2.36E+01 8.73E+05 8.02E-01 2.97E+04 0.00E+00 0.00E+00 I-135 1.14E+02 4.21E+06 3.87E+00 1.43E+05 0.00E+00 0.00E+00 Rb-88 1.07E+07 3.96E+11 3.64E+05 1.35E+10 1.95E-03 7.21E+01 Cs-137 6.41E+05 2.37E+10 2.18E+04 8.07E+08 0.00E+00 0.00E+00 Ba-137m 6.41E+05 2.37E+10 2.18E+04 8.07E+08 0.00E+00 0.00E+00 12.2-55 Rev. 2

APR1400 DCD TIER 2 Table 12.2-20 (1 of 3)

Liquid Radwaste System Tank Source Terms (Bq/cm3)

Equipment Floor Chemical Nuclide Waste Tank Drain Tank Waste Tank Monitor Tank Br-84 6.27E+01 8.62E+01 1.96E+00 2.00E-03 I-131 7.94E+03 1.09E+04 2.48E+02 2.17E+01 I-132 2.13E+03 2.93E+03 6.66E+01 4.33E-01 I-133 1.13E+04 1.55E+04 3.52E+02 1.30E+01 I-134 1.30E+03 1.79E+03 4.07E+01 6.78E-02 I-135 6.40E+03 8.80E+03 2.00E+02 2.58E+00 Rb-88 7.33E+03 1.01E+04 2.29E+02 6.53E-01 Cs-134 1.13E+03 1.55E+03 3.52E+01 1.76E+01 Cs-136 1.54E+02 2.12E+02 4.81E+00 2.21E+00 Cs-137 1.30E+03 1.79E+03 4.07E+01 2.03E+01 Na-24 5.79E+02 7.96E+02 1.81E+01 5.07E-02 Cr-51 1.75E+02 2.41E+02 5.48E+00 5.27E-02 Mn-54 2.03E+01 2.79E+01 6.34E-01 6.31E-03 Fe-55 1.52E+01 2.09E+01 4.75E-01 4.74E-03 Fe-59 3.81E+00 5.24E+00 1.19E-01 1.16E-03 Co-58 5.82E+01 8.01E+01 1.82E+00 1.79E-02 Co-60 6.72E+00 9.24E+00 2.10E-01 2.10E-03 Zn-65 6.46E+00 8.89E+00 2.02E-01 2.01E-03 12.2-56 Rev. 2

APR1400 DCD TIER 2 Table 12.2-20 (2 of 3)

Equipment Floor Chemical Nuclide Waste Tank Drain Tank Waste Tank Monitor Tank Sr-89 1.04E+01 1.43E+01 3.26E-01 3.19E-03 Sr-90 6.98E-01 9.59E-01 2.18E-02 2.18E-04 Sr-91 1.54E+01 2.12E+01 4.81E-01 8.91E-04 Y-91m 8.99E+00 1.24E+01 2.81E-01 5.67E-04 Y-91 1.54E+00 2.12E+00 4.81E-02 5.00E-04 Y-93 3.68E-01 5.06E-01 1.15E-02 2.24E-05 Zr-95 4.93E+00 6.78E+00 1.54E-01 1.51E-03 Nb-95 1.66E+00 2.28E+00 5.18E-02 5.47E-04 Mo-99 8.99E+02 1.24E+03 2.81E+01 1.95E-01 Tc-99m 5.22E+02 7.17E+02 1.63E+01 1.75E-01 Ru-103 5.57E-01 7.66E-01 1.74E-02 1.69E-04 Ru-106 2.37E-01 3.26E-01 7.40E-03 7.37E-04 Ag-110m 1.65E+01 2.27E+01 5.15E-01 5.12E-03 Te-129m 1.89E+01 2.60E+01 5.92E-01 5.73E-03 Te-129 2.01E+01 2.77E+01 6.29E-01 3.73E-03 Te-131m 8.99E+01 1.24E+02 2.81E+00 1.33E-02 Te-131 3.55E+01 4.88E+01 1.11E+00 2.48E-03 Te-132 6.27E+02 8.62E+02 1.96E+01 1.43E-01 Ba-137m 1.18E+03 1.63E+03 3.70E+01 1.90E+01 Ba-140 1.30E+01 1.79E+01 4.07E-01 3.74E-03 12.2-57 Rev. 2

APR1400 DCD TIER 2 Table 12.2-20 (3 of 3)

Equipment Floor Chemical Nuclide Waste Tank Drain Tank Waste Tank Monitor Tank La-140 4.38E+00 6.03E+00 1.37E-01 2.45E-03 Ce-141 4.86E-01 6.69E-01 1.52E-02 1.47E-04 Ce-143 1.30E+00 1.79E+00 4.07E-02 2.05E-04 Ce-144 1.42E+00 1.95E+00 4.44E-02 4.42E-04 W-187 3.10E+01 4.27E+01 9.70E-01 3.95E-03 Np-239 2.76E+01 3.79E+01 8.62E-01 5.63E-03 12.2-58 Rev. 2

APR1400 DCD TIER 2 Table 12.2-21 (1 of 3)

Liquid Radwaste System Component Source Terms (Bq)

Nuclide Reverse Osmosis Cation Bed Mixed Bed 1 Mixed Bed 2 Na-24 2.14E+11 2.15E+10 2.36E+09 2.15E+07 Cr-51 5.58E+11 5.58E+10 6.14E+09 5.58E+07 Mn-54 7.11E+10 7.11E+09 7.82E+08 7.11E+06 Fe-55 5.36E+10 5.36E+09 5.90E+08 5.36E+06 Co-58 1.98E+11 1.98E+10 2.17E+09 1.98E+07 Fe-59 1.26E+10 1.26E+09 1.39E+08 1.26E+06 Co-60 2.37E+10 2.37E+09 2.61E+08 2.37E+06 Zn-65 2.26E+10 2.26E+09 2.49E+08 2.26E+06 Br-84 8.79E+08 1.23E+07 1.02E+08 9.40E+05 Rb-88 6.06E+10 6.78E+09 4.41E+08 3.39E+08 Sr-89 3.49E+10 3.49E+09 3.84E+08 3.49E+06 Y-89m 3.49E+06 3.49E+05 3.83E+04 3.49E+02 Sr-90 2.47E+09 2.47E+08 2.71E+07 2.47E+05 Y-90m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Y-90 1.49E+09 1.49E+08 1.64E+07 1.49E+05 Sr-91 3.67E+09 3.69E+08 4.05E+07 3.69E+05 Y-91m 2.34E+09 2.35E+08 2.58E+07 2.35E+05 Y-91 5.52E+09 5.53E+08 6.08E+07 5.53E+05 Y-93 9.27E+07 9.30E+06 1.02E+06 9.30E+03 Zr-93 9.40E-01 9.40E-02 1.03E-02 9.40E-05 Nb-93m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Zr-95 1.67E+10 1.67E+09 1.83E+08 1.67E+06 Nb-95m 1.69E+08 1.69E+07 1.86E+06 1.69E+04 Nb-95 6.74E+09 6.74E+08 7.42E+07 6.74E+05 Mo-99 1.31E+12 1.31E+11 1.44E+10 1.31E+08 Tc-99m 1.20E+12 1.21E+11 1.33E+10 1.21E+08 Tc-99 6.99E+04 6.99E+03 7.68E+02 6.99E+00 Ru-103 1.83E+09 1.83E+08 2.01E+07 1.83E+05 12.2-59 Rev. 2

APR1400 DCD TIER 2 Table 12.2-21 (2 of 3)

Nuclide Reverse Osmosis Cation Bed Mixed Bed 1 Mixed Bed 2 Rh-103m 1.82E+09 1.82E+08 2.00E+07 1.82E+05 Ru-106 8.31E+09 8.31E+08 9.14E+07 8.31E+05 Rh-106m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Rh-106 8.31E+09 8.31E+08 9.14E+07 8.31E+05 Ag-110m 5.76E+10 5.76E+09 6.34E+08 5.76E+06 Ag-110 7.49E+08 7.49E+07 8.24E+06 7.49E+04 Te-129m 6.15E+10 6.15E+09 6.76E+08 6.15E+06 Te-129 3.96E+10 3.96E+09 4.36E+08 3.96E+06 I-129 3.08E+01 3.08E+00 3.38E-01 3.08E-03 Te-131m 6.58E+10 6.59E+09 7.24E+08 6.59E+06 Te-131 1.22E+10 1.23E+09 1.35E+08 1.23E+06 I-131 1.98E+13 4.49E+09 2.18E+12 1.98E+10 Xe-131m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Te-132 1.01E+12 1.01E+11 1.12E+10 1.01E+08 I-132 1.13E+12 1.01E+11 2.47E+10 2.25E+08 I-133 5.82E+12 2.21E+09 6.41E+11 5.83E+09 Xe-133m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe-133 0.00E+00 0.00E+00 0.00E+00 0.00E+00 I-134 2.90E+10 2.56E+08 3.31E+09 3.03E+07 Cs-134 3.97E+12 3.97E+11 2.21E+10 1.99E+10 I-135 1.06E+12 1.26E+09 1.17E+11 1.07E+09 Xe-135m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe-135 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Cs-135 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Cs-136 4.36E+11 4.36E+10 2.42E+09 2.18E+09 Cs-137 4.61E+12 4.61E+11 2.56E+10 2.30E+10 Ba-137m 4.31E+12 4.31E+11 2.40E+10 2.15E+10 Ba-140 3.68E+10 3.68E+09 4.05E+08 3.68E+06 12.2-60 Rev. 2

APR1400 DCD TIER 2 Table 12.2-21 (3 of 3)

Nuclide Reverse Osmosis Cation Bed Mixed Bed 1 Mixed Bed 2 La-140 3.21E+10 3.21E+09 3.54E+08 3.21E+06 Ce-141 1.57E+09 1.57E+08 1.73E+07 1.57E+05 Ce-143 1.04E+09 1.05E+08 1.15E+07 1.05E+05 Pr-143 3.02E+08 3.02E+07 3.32E+06 3.02E+04 Ce-144 4.97E+09 4.98E+08 5.47E+07 4.98E+05 Pr-144 4.96E+09 4.96E+08 5.46E+07 4.96E+05 W-187 1.82E+10 1.82E+09 2.00E+08 1.82E+06 Np-239 3.53E+10 3.53E+09 3.89E+08 3.53E+06 12.2-61 Rev. 2

APR1400 DCD TIER 2 Table 12.2-22 Solid Radwaste System Tank Source Terms (Bq)

Spent Resin Low-Activity Spent Resin Low-Activity Long-Term Spent Resin Long-Term Spent Resin Nuclide Storage Tank (1) Tank (2) Nuclide Storage Tank Tank Na-24 7.01E+11 7.16E+10 Rh-106 0.00E+00 2.77E+09 Cr-51 9.49E+12 1.86E+11 Ag-110m 9.29E+12 1.92E+10 Mn-54 1.45E+13 2.37E+10 Ag-110 0.00E+00 2.50E+08 Fe-55 3.40E+13 1.79E+10 Te-129m 1.31E+12 2.05E+10 Co-58 8.28E+12 6.60E+10 Te-129 1.92E+09 1.32E+10 Fe-59 3.34E+11 4.20E+09 I-129 0.00E+00 1.03E+01 Co-60 2.38E+13 7.90E+09 Te-131m 2.12E+11 2.20E+10 Zn-65 3.57E+12 7.53E+09 Te-131 1.21E+09 4.10E+09 Br-84 2.73E+09 3.46E+08 I-131 4.04E+11 6.61E+12 Rb-88 1.15E+11 2.27E+10 Te-132 4.04E+12 3.37E+11 Sr-89 1.01E+12 1.16E+10 I-132 4.04E+11 3.78E+11 Y-89m 0.00E+00 1.16E+06 I-133 1.92E+13 1.95E+12 Sr-90 3.96E+12 8.23E+08 I-134 9.30E+10 1.08E+10 Y-90 0.00E+00 4.97E+08 Cs-134 1.20E+15 1.32E+12 Sr-91 1.20E+10 1.23E+09 I-135 3.54E+12 3.58E+11 Y-91m 9.40E+07 7.83E+08 Cs-136 2.34E+12 1.45E+11 Y-91 1.63E+07 1.84E+09 Cs-137 4.60E+15 1.53E+12 Y-93 3.80E+06 3.10E+07 Ba-137m 4.60E+15 1.43E+12 Zr-93 0.00E+00 3.13E-01 Ba-140 3.21E+11 1.23E+10 Zr-95 6.34E+11 5.56E+09 La-140 1.40E+10 1.07E+10 Nb-95m 0.00E+00 5.63E+07 Ce-141 3.11E+10 5.23E+08 Nb-95 1.10E+11 2.25E+09 Ce-143 3.51E+09 3.50E+08 Mo-99 4.81E+12 4.37E+11 Pr-143 0.00E+00 1.01E+08 Tc-99m 2.50E+11 4.03E+11 Ce-144 9.18E+11 1.66E+09 Tc-99 0.00E+00 2.33E+04 Pr-144 0.00E+00 1.65E+09 Ru-103 4.32E+10 6.10E+08 W-187 6.01E+10 6.07E+09 Rh-103m 0.00E+00 6.07E+08 Np-239 1.30E+11 1.18E+10 Ru-106 2.01E+11 2.77E+09 (1) Source terms for the spent resin long-term storage tank are calculated based on a maximum of 10 years of cumulative radioactive resin batches from the CVCS demineralizers with decay.

(2) Source terms for the low-activity spent resin storage tank are calculated based on the using the source term for LWMS spent resin and multiplying by a factor of 3 for conservative shielding design.

12.2-62 Rev. 2

APR1400 DCD TIER 2 Table 12.2-22a Waste Drum Storage Area Source Terms (Bq)

Nuclide Activity Nuclide Activity I-131 2.96E+12 Y-91 5.35E+09 I-132 6.15E-82 Y-93 6.34E-13 I-133 2.12E+03 Nb-95 1.42E+10 I-134 3.77E-235 Mo-99 2.33E+09 I-135 5.15E-20 Tc-99m 2.57E-24 Cs-134 5.47E+12 Ru-103 1.63E+09 Cs-136 1.57E+11 Ru-106 1.08E+10 Cs-137 6.56E+12 Ag-110m 7.64E+10 Cr-51 4.11E+11 Sb-125 5.30E+10 Mn-54 9.70E+10 Te-129m 1.42E-176 Fe-55 7.38E+10 Te-129 2.66E+04 Fe-59 1.19E+10 Te-131m 0.00E+00 Co-58 2.55E+11 Te-132 5.26E+09 Co-60 2.31E+11 Ba-137m 5.86E+12 Zr-95 2.05E+10 Ba-140 1.27E+10 Zn-65 2.95E+10 La-140 8.88E+04 Na-24 1.02E-02 Ce-141 1.27E+09 Sr-89 3.43E+10 Ce-143 1.75E+03 Sr-90 3.45E+09 Ce-144 6.55E+09 Sr-91 1.17E-12 W-187 1.24E+02 Y-91m 1.10E-251 Np-239 2.00E+07 12.2-63 Rev. 2

APR1400 DCD TIER 2 Table 12.2-23 (1 of 4)

Airborne Radioactivity Concentrations Reactor Containment Building (Normal Operation)

Airborne 10 CFR 20 Airborne 10 CFR 20 Concentration (1) Appendix B DAC Concentration Appendix B DAC Nuclide (Bq/cm3) (Bq/cm3) Fraction Nuclide (Bq/cm3) (Bq/cm3) Fraction Kr-85m 8.60E-02 7.40E-01 1.16E-01 Zn-65 9.83E-02 3.70E-03 2.66E+01 Kr-85 7.13E+02 3.70E+00 1.93E+02 Sr-89 4.15E-02 2.22E-03 1.87E+01 Kr-87 1.92E-02 1.85E-01 1.04E-01 Sr-90 2.05E-02 7.40E-05 2.77E+02 Kr-88 1.19E-01 7.40E-02 1.61E+00 Sr-91 4.80E-04 3.70E-02 1.30E-02 Xe-131m 2.34E+01 1.48E+01 1.58E+00 Y-91m 2.45E-05 2.59E+00 9.45E-06 Xe-133m 2.62E-01 3.70E+00 7.09E-02 Y-91 7.09E-03 1.85E-03 3.83E+00 Xe-133 6.73E+02 3.70E+00 1.82E+02 Y-93 1.22E-05 3.70E-02 3.30E-04 Xe-135m 2.79E-03 3.33E-01 8.38E-03 Zr-95 2.48E-02 1.85E-03 1.34E+01 Xe-135 9.95E-01 3.70E-01 2.69E+00 Nb-95 4.60E-03 1.85E-02 2.48E-01 Xe-138 2.26E-03 1.48E-01 1.53E-02 Mo-99 1.95E-01 2.22E-02 8.79E+00 Br-84 1.09E-04 7.40E-01 1.47E-04 Tc-99m 1.03E-02 2.22E+00 4.66E-03 I-131 5.04E+00 7.40E-04 6.81E+03 Ru-103 1.73E-03 1.11E-02 1.56E-01 I-132 1.61E-02 1.11E-01 1.45E-01 Ru-106 4.43E-03 1.85E-04 2.39E+01 I-133 7.71E-01 3.70E-03 2.08E+02 Ag-110m 2.53E-01 1.48E-03 1.71E+02 I-134 3.75E-03 7.40E-01 5.07E-03 Te-129m 5.02E-02 3.70E-03 1.36E+01 I-135 1.39E-01 2.59E-02 5.36E+00 Te-129 7.67E-05 1.11E+00 6.91E-05 Rb-88 7.13E-03 1.11E+00 6.43E-03 Te-131m 8.87E-03 7.40E-03 1.20E+00 Cs-134 2.65E+01 1.48E-03 1.79E+04 Te-131 4.88E-05 7.40E-02 6.59E-04 Cs-136 1.59E-01 1.11E-02 1.44E+01 Te-132 1.61E-01 3.33E-03 4.84E+01 Cs-137 3.83E+01 2.22E-03 1.73E+04 Ba-140 1.31E-02 2.22E-02 5.92E-01 Na-24 2.86E-02 7.40E-02 3.86E-01 La-140 5.81E-04 1.85E-02 3.14E-02 Cr-51 3.84E-01 2.96E-01 1.30E+00 Ce-141 1.25E-03 7.40E-03 1.69E-01 Mn-54 3.52E-01 1.11E-02 3.17E+01 Ce-143 1.41E-04 2.59E-02 5.46E-03 Fe-55 3.75E-01 2.96E-02 1.27E+01 Ce-144 2.34E-02 2.22E-04 1.06E+02 Fe-59 1.35E-02 3.70E-03 3.65E+00 W-187 2.44E-03 1.48E-01 1.65E-02 Co-58 3.24E-01 1.11E-02 2.92E+01 Np-239 5.44E-03 3.33E-02 1.63E-01 Co-60 1.83E-01 3.70E-04 4.94E+02 H-3 7.92E-01 7.40E-01 1.07E+00 Sum of DAC fractions 4.39E+04 (1) The airborne activity for other nuclides (primarily particulates) in the containment building is calculated based on the calculation of flashing fraction.

12.2-64 Rev. 2

APR1400 DCD TIER 2 Table 12.2-23 (2 of 4)

Reactor Containment Building (48 hr after Shutdown)

Airborne 10 CFR 20 Airborne 10 CFR 20 Concentration Appendix B DAC Concentration Appendix B DAC Nuclide (Bq/cm3) (Bq/cm3) Fraction Nuclide (Bq/cm3) (Bq/cm3) Fraction Kr-85m 1.41E-06 7.40E-01 1.90E-06 Zn-65 4.05E-07 3.70E-03 1.10E-04 Kr-85 8.95E-03 3.70E+00 2.42E-03 Sr-89 6.40E-07 2.22E-03 2.88E-04 Kr-87 1.14E-14 1.85E-01 6.14E-14 Sr-90 4.40E-08 7.40E-05 5.94E-04 Kr-88 4.60E-08 7.40E-02 6.21E-07 Sr-91 3.10E-08 3.70E-02 8.39E-07 Xe-131m 7.98E-03 1.48E+01 5.40E-04 Y-91m 4.12E-24 2.59E+00 1.59E-24 Xe-133m 2.92E-04 3.70E+00 7.90E-05 Y-91 9.48E-08 1.85E-03 5.12E-05 Xe-133 4.50E-01 3.70E+00 1.22E-01 Y-93 9.14E-10 3.70E-02 2.47E-08 Xe-135m 0.00E+00 3.33E-01 0.00E+00 Zr-95 3.04E-07 1.85E-03 1.64E-04 Xe-135 3.27E-04 3.70E-01 8.84E-04 Nb-95 1.00E-07 1.85E-02 5.43E-06 Xe-138 0.00E+00 1.48E-01 0.00E+00 Mo-99 3.45E-05 2.22E-02 1.56E-03 Br-84 0.00E+00 7.40E-01 0.00E+00 Tc-99m 1.45E-07 2.22E+00 6.54E-08 I-131 3.43E-05 7.40E-04 4.63E-02 Ru-103 3.39E-08 1.11E-02 3.05E-06 I-132 7.49E-12 1.11E-01 6.75E-11 Ru-106 1.49E-08 1.85E-04 8.04E-05 I-133 1.20E-05 3.70E-03 3.24E-03 Ag-110m 1.03E-06 1.48E-03 6.98E-04 I-134 4.24E-22 7.40E-01 5.73E-22 Te-129m 1.15E-06 3.70E-03 3.10E-04 I-135 2.32E-07 2.59E-02 8.95E-06 Te-129 7.22E-19 1.11E+00 6.50E-19 Rb-88 0.00E+00 1.11E+00 0.00E+00 Te-131m 1.91E-06 7.40E-03 2.58E-04 Cs-134 7.09E-05 1.48E-03 4.79E-02 Te-131 0.00E+00 7.40E-02 0.00E+00 Cs-136 8.74E-06 1.11E-02 7.88E-04 Te-132 2.60E-05 3.33E-03 7.82E-03 Cs-137 8.21E-05 2.22E-03 3.70E-02 Ba-140 7.38E-07 2.22E-02 3.32E-05 Na-24 4.13E-06 7.40E-02 5.59E-05 La-140 1.23E-07 1.85E-02 6.64E-06 Cr-51 1.05E-05 2.96E-01 3.56E-05 Ce-141 2.94E-08 7.40E-03 3.97E-06 Mn-54 1.27E-06 1.11E-02 1.15E-04 Ce-143 3.05E-08 2.59E-02 1.18E-06 Fe-55 9.56E-07 2.96E-02 3.23E-05 Ce-144 8.91E-08 2.22E-04 4.01E-04 Fe-59 2.33E-07 3.70E-03 6.29E-05 W-187 4.98E-07 1.48E-01 3.37E-06 Co-58 3.60E-06 1.11E-02 3.24E-04 Np-239 1.04E-06 3.33E-02 3.11E-05 Co-60 4.23E-07 3.70E-04 1.14E-03 H-3 1.74E-06 7.40E-01 2.35E-06 Sum of DAC fractions 2.75E-01 12.2-65 Rev. 2

APR1400 DCD TIER 2 Table 12.2-23 (3 of 4)

Auxiliary Building Cubicles (Normal Operation)

Airborne Radioactivity Derived Air Concentration (DAC) Fraction Concentration (Bq/cm3)

Cubicle Kr, Xe Br, I H-3 Others Kr, Xe Br, I H-3 Others Total CS Pump and Miniflow HX Rm (050-A01C,D) 7.37E-02 4.92E-07 9.82E-03 2.86E-06 1.99E-02 5.18E-04 1.33E-02 1.40E-03 3.50E-02 SI Pump Rm (050-A02C,D) 1.16E-01 7.81E-07 1.55E-02 4.54E-06 3.13E-02 8.18E-04 2.09E-02 2.19E-03 5.52E-02 Floor Drain Sump Pump Rm 1.84E+00 5.34E-05 4.10E-02 1.05E-04 7.99E-01 2.61E-02 5.54E-02 1.55E-02 8.96E-01 (055-A34A,B,C,D)

Pipe Chase and Valve Rm (055-A14C) 9.44E-02 6.20E-07 1.27E-02 3.75E-06 2.54E-02 6.64E-04 1.72E-02 1.82E-03 4.50E-02 Shutdown Cooling HX Rm (055-A30A,B) 3.63E-02 2.39E-07 4.88E-03 1.44E-06 9.77E-03 2.55E-04 6.59E-03 6.98E-04 1.73E-02 Charging Pump Rm (055-A55B) 1.58E+00 2.68E-07 3.56E-02 1.12E-05 6.44E-01 1.39E-04 4.81E-02 3.47E-03 6.95E-01 Charging Pump Miniflow HX Rm (055-A43 A) 4.49E-01 1.23E-05 1.01E-02 3.07E-06 1.81E-01 6.37E-03 1.37E-02 9.87E-04 2.02E-01 Equipment Drain Tank Rm (055-A51B) 8.48E-02 1.77E-06 3.59E-04 2.53E-06 3.62E-02 1.03E-03 4.85E-04 7.14E-04 3.84E-02 Reactor Drain Pump Rm (055-A52B,A53B) 2.70E+00 8.76E-06 2.11E-03 1.75E-05 7.95E-01 5.20E-03 2.85E-03 3.93E-03 8.07E-01 Gas Stripper Rm (068-A06A) 2.14E+00 6.11E-07 3.53E-07 9.02E-01 3.04E-04 4.76E-07 9.02E-01 Filter and Demin. Valve Area (068-A10A) 1.90E+00 4.85E-05 3.67E-02 9.76E-05 7.92E-01 2.37E-02 4.96E-02 1.42E-02 8.79E-01 SFP Cleanup Pump Rm (078-A38A) 4.16E-02 2.87E-07 5.17E-03 1.53E-06 1.12E-02 3.00E-04 6.99E-03 7.33E-04 1.92E-02 Reactor Makeup Water Pump Rm (078-A49A) 2.49E-07 1.36E-12 1.20E-02 1.95E-10 6.57E-08 1.81E-09 1.63E-02 9.88E-08 1.63E-02 Holdup Pump Rm (078-A50B) 5.29E-04 5.88E-08 1.19E-02 1.11E-07 2.95E-04 6.30E-04 1.61E-02 4.05E-05 1.65E-02 Volume Control Tank Rm (100-A25A) 1.06E+00 1.62E-07 2.41E-021 5.66E-06 3.88E-01 9.22E-05 3.26E-02 2.37E-03 4.23E-01 Valve Rm (120-A23A) 1.81E-01 5.27E-06 5.01E-03 1.06E-05 7.89E-02 2.58E-03 6.78E-03 1.73E-03 9.00E-02 Fuel Handling Area (Normal Operation) 2.40E-02 3.20E-02 3.20E-02 Fuel Handling Area (Refueling) 5.70E-02 7.70E-02 7.70E-02 12.2-66 Rev. 2

APR1400 DCD TIER 2 Table 12.2-23 (4 of 4)

Compound Building Cubicles (Normal Operation)

Airborne Radioactivity Derived Air Concentration (DAC) Fraction Concentration (Bq/cm3)

Cubicle Kr, Xe Br, I H-3 Others Kr, Xe Br, I H-3 Others Total Valve Rm (063-P07) 1.62E+00 1.03E-04 8.29E-02 1.55E-04 6.66E-01 5.24E-02 1.12E-01 3.13E-02 8.62E-01 Equipment Waste Pump Rm (063-P21,P22) 1.30E+00 3.72E-05 2.92E-02 6.29E-05 5.52E-01 1.85E-02 3.94E-02 1.10E-02 6.21E-01 Equipment Waste Tank Rm (063-P23,P24) 1.73E-01 4.87E-06 7.33E-04 7.60E-06 7.20E-02 2.46E-03 9.91E-04 1.47E-03 7.69E-02 Floor Drain Pump Rm (063-P25) 2.17E+00 5.93E-05 4.89E-02 8.40E-05 8.71E-01 3.08E-02 6.61E-02 1.84E-02 9.87E-01 Normal Sump Pump Rm (063-P26) 1.90E-01 5.52E-06 4.23E-03 1.09E-05 8.26E-02 2.69E-03 5.72E-03 1.60E-03 9.26E-02 Chemical Waste Pump Rm (063-P27) 5.50E-02 1.46E-06 1.13E-03 2.73E-06 2.18E-02 7.16E-04 1.52E-03 4.26E-04 2.45E-02 Floor Drain Tank Rm (063-P28,P29) 4.74E-01 1.30E-05 1.07E-02 1.88E-05 1.92E-01 6.73E-03 1.44E-02 4.02E-03 2.17E-01 Chemical Waste Tank Rm (063-P30,P31) 1.07E-02 2.83E-07 2.42E-04 3.80E-07 4.15E-03 1.51E-04 3.27E-04 9.13E-05 4.72E-03 Detergent Waste Tank and Pump Rm 0.00E+00 5.85E-08 1.02E-05 3.21E-08 0.00E+00 3.33E-05 1.38E-05 8.59E-06 5.57E-05 (063-P32)

Chemical Drain Sump Pump Rm (063-P36) 1.00E-01 2.92E-06 2.24E-03 5.79E-06 4.37E-02 1.42E-03 3.02E-03 8.44E-04 4.90E-02 Monitor Tank Rm (063-P37) 0.00E+00 1.08E-08 0.00E+00 1.44E-07 0.00E+00 1.00E-05 0.00E+00 7.51E-05 8.52E-05 Monitor Tank Pump Rm (063-P54) 0.00E+00 2.07E-07 0.00E+00 2.58E-06 0.00E+00 1.84E-04 0.00E+00 1.34E-03 1.53E-03 Valve Rm (085-P06) 0.00E+00 1.98E-05 1.62E-02 2.86E-05 0.00E+00 1.02E-02 2.19E-02 6.10E-03 3.81E-02 Valve Rm (085-P15) 0.00E+00 4.61E-05 3.64E-02 7.48E-05 0.00E+00 2.31E-02 4.92E-02 1.37E-02 8.61E-02 Valve Rm (085-P16) 2.06E-01 5.96E-06 4.59E-03 1.08E-05 8.91E-02 2.92E-03 6.21E-03 1.64E-03 9.45E-02 12.2-67 Rev. 2

APR1400 DCD TIER 2 Table 12.2-24 (1 of 14)

Source Terms for Post-Accident Shielding Analysis

a. Initial Release Fraction from the Core Early In-Vessel Release Plus Gap Release Gap Release Source Term(1) Nuclide Group(2) Percentage Nuclide Group(2) Percentage Liquid-containing 2 and 3 5% 2 40 %

systems(3) 3 30 %

4 5%

5 2%

6 0.25 %

7 0.02 %

8 0.05 %

Airborne 1, 2, and 3 5% 1 100 %

2 40 %

3 30 %

4 5%

5 2%

6 0.25 %

7 0.02 %

8 0.05 %

(1) The source terms represent the initial releases from the core into the reactor containment building sump water and atmosphere.

(2) Nuclide Group 1: Xe, Kr Nuclide Group 2: I, Br Nuclide Group 3: Cs, Rb Nuclide Group 4: Te, Sb, Se Nuclide Group 5: Sr, Ba Nuclide Group 6: Co, Ru, Rh, Pd, Mo, Tc Nuclide Group 7: Am, Cm, La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y Nuclide Group 8: Ce, Pu, Np (3) The following components of the engineered safety systems use water in the IRWST during post-accident conditions:

- Shutdown Cooling System : shutdown cooling pumps, shutdown cooling mini flow heat exchanger

- Safety Injection System : safety injection pump

- Containment Spray System : containment spray pump, containment spray heat exchanger, containment spray mini flow heat exchanger

- Piping : associated pipes of the above systems 12.2-68 Rev. 2

APR1400 DCD TIER 2 Table 12.2-24 (2 of 14)

b. Radioactive Concentration of Post-Accident Recirculating Water (Bq/cc) (1 of 5)

Elasped Time after the onset of the Accident Nuclide 0 hr 1 hr 1 day 1 week 1 month Se-84 2.75E+07 9.25E+01 0.00E+00 0.00E+00 0.00E+00 Se-85 1.37E+07 2.24E-21 0.00E+00 0.00E+00 0.00E+00 Se-87 2.03E+07 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Br-84 2.25E+08 6.08E+07 5.26E-06 8.54E-88 0.00E+00 Br-85 2.59E+08 1.30E+02 0.00E+00 0.00E+00 0.00E+00 Br-87 4.25E+08 1.61E-11 0.00E+00 0.00E+00 0.00E+00 Br-88 4.28E+08 1.40E-58 0.00E+00 0.00E+00 0.00E+00 Br-89 2.89E+08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr-85 0.00E+00 3.93E+00 2.73E+01 2.79E+01 2.78E+01 Kr-85m 0.00E+00 2.39E+06 6.80E+04 1.43E-05 1.16E-42 Rb-86 1.59E+06 1.59E+06 1.53E+06 1.23E+06 5.22E+05 Rb-88 5.74E+08 5.55E+07 2.55E-16 0.00E+00 0.00E+00 Rb-89 7.63E+08 5.16E+07 6.41E-20 0.00E+00 0.00E+00 Rb-90 7.20E+08 9.01E+01 0.00E+00 0.00E+00 0.00E+00 Rb-91 9.04E+08 2.17E-10 0.00E+00 0.00E+00 0.00E+00 Sr-89 4.68E+07 4.69E+07 4.63E+07 4.27E+07 3.11E+07 Sr-90 4.21E+06 4.21E+06 4.21E+06 4.21E+06 4.20E+06 Sr-91 6.30E+07 5.86E+07 1.09E+07 2.99E+02 9.65E-16 Sr-92 6.35E+07 4.92E+07 1.37E+05 1.38E-11 6.67E-73 Sr-95 5.93E+07 1.24E-34 0.00E+00 0.00E+00 0.00E+00 Y-90 4.44E+04 8.92E+04 9.97E+05 3.53E+06 4.20E+06 Y-91 5.74E+05 5.92E+05 9.13E+05 9.24E+05 7.03E+05 Y-91m 3.64E+05 1.98E+07 6.88E+06 1.88E+02 6.07E-16 Y-92 6.41E+05 1.04E+07 1.45E+06 1.08E-06 1.23E-53 Y-93 4.60E+05 4.29E+05 8.86E+04 4.52E+00 1.60E-16 Y-95 7.08E+05 1.35E+04 3.68E-36 0.00E+00 0.00E+00 Zr-93 0.00E+00 2.30E-05 2.80E-04 3.46E-04 3.46E-04 Zr-95 6.71E+05 6.71E+05 6.64E+05 6.22E+05 4.85E+05 Zr-97 6.37E+05 6.11E+05 2.38E+05 6.48E+02 9.53E-08 Nb-93m 0.00E+00 6.29E-11 2.30E-08 2.88E-07 1.32E-06 Nb-95 6.68E+05 6.68E+05 6.68E+05 6.65E+05 6.21E+05 12.2-69 Rev. 2

APR1400 DCD TIER 2 Table 12.2-24 (3 of 14)

b. Radioactive Concentration of Post-Accident Recirculating Water (Bq/cc) (2 of 5)

Elasped Time after the onset of the Accident Nuclide 0 hr 1 hr 1 day 1 week 1 month Nb-95m 0.00E+00 4.28E+01 9.33E+02 3.79E+03 4.09E+03 Nb-97 0.00E+00 2.69E+05 2.56E+05 6.98E+02 1.03E-07 Nb-97m 0.00E+00 5.80E+05 2.26E+05 6.14E+02 9.04E-08 Nb-99 4.14E+05 6.81E-71 0.00E+00 0.00E+00 0.00E+00 Mo-99 8.74E+06 8.65E+06 6.79E+06 1.50E+06 4.56E+03 Mo-103 7.79E+06 6.76E-12 0.00E+00 0.00E+00 0.00E+00 Tc-99 0.00E+00 3.23E-03 7.25E-02 2.76E-01 3.34E-01 Tc-99m 7.70E+06 7.70E+06 6.57E+06 1.46E+06 4.44E+03 Tc-103 7.89E+06 1.67E-15 0.00E+00 0.00E+00 0.00E+00 Tc-106 4.52E+06 2.32E-23 0.00E+00 0.00E+00 0.00E+00 Ru-103 7.94E+06 7.93E+06 7.80E+06 7.02E+06 4.68E+06 Ru-105 6.04E+06 5.17E+06 1.43E+05 2.46E-05 9.23E-43 Ru-106 3.57E+06 3.57E+06 3.56E+06 3.52E+06 3.37E+06 Rh-103m 0.00E+00 4.14E+06 7.79E+06 7.01E+06 4.67E+06 Rh-105 5.40E+06 5.40E+06 3.89E+06 2.33E+05 4.65E+00 Rh-105m 0.00E+00 1.27E+06 3.50E+04 6.04E-06 2.27E-43 Rh-106 0.00E+00 3.57E+06 3.56E+06 3.52E+06 3.37E+06 Sb-127 8.86E+06 8.79E+06 7.40E+06 2.51E+06 3.99E+04 Sb-129 3.03E+07 2.59E+07 6.91E+05 9.72E-05 1.67E-42 Sb-131 7.44E+07 1.22E+07 1.06E-11 0.00E+00 0.00E+00 Sb-132 4.33E+07 1.53E+01 0.00E+00 0.00E+00 0.00E+00 Sb-133 6.85E+07 2.04E+00 0.00E+00 0.00E+00 0.00E+00 Sb-134 1.32E+07 4.00E-92 0.00E+00 0.00E+00 0.00E+00 Te-127 8.77E+06 8.77E+06 8.14E+06 3.75E+06 1.29E+06 Te-127m 1.48E+06 1.48E+06 1.48E+06 1.45E+06 1.27E+06 Te-129 2.89E+07 2.80E+07 4.44E+06 3.23E+06 2.01E+06 Te-129m 5.90E+06 5.90E+06 5.81E+06 5.13E+06 3.19E+06 Te-131 7.56E+07 1.77E+07 2.50E+06 8.96E+04 2.59E-01 Te-131m 1.93E+07 1.89E+07 1.11E+07 3.98E+05 1.15E+00 Te-132 1.29E+08 1.28E+08 1.04E+08 2.91E+07 2.18E+05 Te-133 1.09E+08 1.05E+07 2.29E-01 2.58E-48 0.00E+00 12.2-70 Rev. 2

APR1400 DCD TIER 2 Table 12.2-24 (4 of 14)

b. Radioactive Concentration of Post-Accident Recirculating Water (Bq/cc) (3 of 5)

Elasped Time after the onset of the Accident Nuclide 0 hr 1 hr 1 day 1 week 1 month Te-133m 9.10E+07 4.30E+07 1.36E+00 1.54E-47 0.00E+00 Te-134 1.92E+08 7.10E+07 8.18E-03 4.90E-65 0.00E+00 Te-135 9.49E+07 3.42E-49 0.00E+00 0.00E+00 0.00E+00 I-129 3.71E+01 3.71E+01 3.71E+01 3.71E+01 3.71E+01 I-131 7.26E+08 7.24E+08 6.67E+08 3.99E+08 5.49E+07 I-132 1.05E+09 8.10E+08 1.08E+08 3.00E+07 2.25E+05 I-133 1.54E+09 1.49E+09 6.95E+08 5.72E+06 5.87E-02 I-134 1.75E+09 8.56E+08 1.43E+01 5.12E-49 0.00E+00 I-135 1.45E+09 1.31E+09 1.17E+08 3.24E+01 2.35E-24 I-137 7.48E+08 6.62E-36 0.00E+00 0.00E+00 0.00E+00 I-138 3.76E+08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe-131m 0.00E+00 1.92E+04 4.30E+05 1.96E+06 1.63E+06 Xe-133 0.00E+00 8.09E+06 1.27E+08 1.20E+08 5.92E+06 Xe-133m 0.00E+00 5.72E+05 8.14E+06 3.07E+06 2.19E+03 Xe-135 0.00E+00 9.53E+07 3.09E+08 1.07E+04 6.22E-15 Xe-135m 0.00E+00 2.08E+08 2.01E+07 5.56E+00 4.04E-25 Cs-134 1.70E+08 1.70E+08 1.70E+08 1.69E+08 1.65E+08 Cs-135 3.65E+02 3.65E+02 3.65E+02 3.65E+02 3.65E+02 Cs-136 4.37E+07 4.36E+07 4.15E+07 3.02E+07 9.00E+06 Cs-137 9.41E+07 9.41E+07 9.41E+07 9.41E+07 9.39E+07 Cs-138 1.12E+09 3.08E+08 3.86E-05 6.51E-86 0.00E+00 Cs-140 9.69E+08 1.00E-08 0.00E+00 0.00E+00 0.00E+00 Ba-137m 5.99E+06 8.90E+07 8.90E+07 8.90E+07 8.89E+07 Ba-139 7.19E+07 4.36E+07 4.37E+02 2.20E-29 0.00E+00 Ba-140 7.12E+07 7.10E+07 6.74E+07 4.87E+07 1.40E+07 Ba-143 5.80E+07 1.20E-72 0.00E+00 0.00E+00 0.00E+00 Ba-144 4.84E+07 1.47E-91 0.00E+00 0.00E+00 0.00E+00 La-140 7.13E+05 1.92E+06 2.39E+07 5.16E+07 1. 61E+07 La-141 6.56E+05 5.50E+05 9.62E+03 9.57E-08 6.40E-50 La-142 6.57E+05 4.25E+05 1.88E+01 1.02E-26 0.00E+00 La-143 6.51E+05 3.34E+04 7.09E-26 0.00E+00 0.00E+00 12.2-71 Rev. 2

APR1400 DCD TIER 2 Table 12.2-24 (5 of 14)

b. Radioactive Concentration of Post-Accident Recirculating Water (Bq/cc) (4 of 5)

Elasped Time after the onset of the Accident Nuclide 0 hr 1 hr 1 day 1 week 1 month La-144 5.93E+05 4.79E-22 0.00E+00 0.00E+00 0.00E+00 Ce-141 1.60E+06 1.60E+06 1.57E+06 1.38E+06 8.46E+05 Ce-143 1.64E+06 1.61E+06 9.91E+05 4.81E+04 4.44E-01 Ce-144 1.16E+06 1.16E+06 1.16E+06 1.14E+06 1.08E+06 Pr-143 6.40E+05 6.42E+05 6.72E+05 5.71E+05 1.78E+05 Pr-144 4.67E+05 1.10E+06 1.16E+06 1.14E+06 1.08E+06 Pr-144m 0.00E+00 1.65E+04 1.65E+04 1.63E+04 1.54E+04 Nd-147 2.57E+05 2.56E+05 2.41E+05 1.65E+05 3.87E+04 Pm-147 0.00E+00 7.74E+00 1.80E+02 1.05E+03 2.47E+03 Sm-147 0.00E+00 2.89E-15 1.63E-12 7.06E-11 8.65E-10 Tl-207 0.00E+00 1.24E-21 1.62E-21 5.24E-22 8.85E-22 Tl-208 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.29E-21 Tl-209 0.00E+00 0.00E+00 3.68E-20 3.20E-20 6.69E-21 Pb-209 0.00E+00 8.72E-21 7.73E-21 6.83E-19 4.82E-19 Pb-210 0.00E+00 0.00E+00 4.00E-16 4.63E-16 0.00E+00 Pb-211 0.00E+00 0.00E+00 1.41E-22 0.00E+00 1.61E-22 Pb-212 0.00E+00 3.13E-22 0.00E+00 0.00E+00 9.82E-21 Pb-214 0.00E+00 1.09E-15 1.09E-15 1.81E-15 4.64E-15 Bi-210 0.00E+00 7.83E-17 4.59E-16 6.85E-16 4.93E-16 Bi-211 0.00E+00 1.76E-21 1.70E-20 0.00E+00 1.26E-21 Bi-213 0.00E+00 2.75E-22 1.03E-18 4.56E-18 2.52E-18 Bi-214 0.00E+00 1.15E-15 1.21E-15 1.55E-15 4.68E-15 Po-210 0.00E+00 5.96E-16 9.25E-16 1.37E-15 1.04E-15 Po-211 0.00E+00 1.52E-24 2.40E-24 0.00E+00 1.61E-25 Po-213 0.00E+00 2.69E-22 1.01E-18 4.46E-18 2.47E-18 Po-214 0.00E+00 3.85E-16 4.37E-16 7.68E-16 3.90E-15 Po-215 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.25E-21 Po-216 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.26E-20 Po-218 0.00E+00 6.35E-16 1.00E-15 1.29E-15 4.64E-15 At-217 0.00E+00 2.15E-18 2.87E-18 4.35E-19 1.32E-19 Rn-220 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.27E-20 12.2-72 Rev. 2

APR1400 DCD TIER 2 Table 12.2-24 (6 of 14)

b. Radioactive Concentration of Post-Accident Recirculating Water (Bq/cc) (5 of 5)

Elasped Time after the onset of the Accident Nuclide 0 hr 1 hr 1 day 1 week 1 month Rn-222 0.00E+00 6.01E-16 2.54E-16 4.20E-16 4.01E-15 Fr-221 0.00E+00 0.00E+00 1.23E-17 4.09E-18 7.08E-19 Fr-223 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.54E-22 Ra-224 0.00E+00 4.69E-22 0.00E+00 5.97E-22 1.33E-21 Ra-225 0.00E+00 3.85E-18 2.41E-18 0.00E+00 2. 46E-18 Ra-226 0.00E+00 3.50E-16 7.44E-16 1.36E-15 6.78E-15 Ra-228 0.00E+00 7.14E-22 0.00E+00 0.00E+00 6.90E-20 Ac-225 0.00E+00 1.84E-18 1.09E-20 3.43E-19 1.95E-18 Ac-227 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.79E-20 Ac-228 0.00E+00 3.55E-20 0.00E+00 2.21E-22 4.05E-20 Th-228 0.00E+00 1.03E-20 0.00E+00 4.76E-20 6.32E-20 Th-229 0.00E+00 2.41E-18 1.83E-18 6.54E-18 4.91E-18 Th-230 0.00E+00 9.13E-16 5.72E-13 2.80E-11 5.14E-10 Th-231 0.00E+00 4.34E-13 2.06E-10 4.29E-09 2.24E-08 Th-232 0.00E+00 1.12E-21 0.00E+00 1.41E-19 2.69E-18 Pa-231 0.00E+00 0.00E+00 0.00E+00 7.13E-16 1.85E-14 Pa-233 0.00E+00 1.29E-12 7.41E-10 3.56E-08 6.00E-07 U-233 0.00E+00 0.00E+00 6.16E-18 1.22E-15 7.35E-14 U-234 0.00E+00 1.93E-06 4.63E-05 3.24E-04 1.39E-03 U-235 0.00E+00 3.22E-11 7.76E-10 5.48E-09 2.36E-08 U-236 0.00E+00 1.82E-09 4.37E-08 3.06E-07 1.31E-06 U-237 0.00E+00 1.59E-02 3.63E-01 1.91E+00 3.54E+00 Np-237 0.00E+00 2.41E-09 5.84E-08 4.26E-07 2.07E-06 Np-239 2.11E+07 2.08E+07 1.57E+07 2.69E+06 3.09E+03 Pu-238 5.97E+03 5.97E+03 5.97E+03 5.97E+03 5.99E+03 Pu-239 2.87E+02 2.87E+02 2.88E+02 2.92E+02 2.93E+02 Pu-240 5.39E+02 5.39E+02 5.39E+02 5.39E+02 5.39E+02 Pu-241 1.52E+05 1.52E+05 1.52E+05 1.52E+05 1.51E+05 Am-241 6.53E+01 6.53E+01 6.60E+01 7.00E+01 8.53E+01 Cm-242 3.27E+04 3.27E+04 3.26E+04 3.17E+04 2.88E+04 Cm-244 4.66E+03 4.66E+03 4.66E+03 4.66E+03 4.65E+03 12.2-73 Rev. 2

APR1400 DCD TIER 2 Table 12.2-24 (7 of 14)

c. Radioactive Concentration of Post-Accident Airborne in Containment (Bq/cc) (1 of 6)

Elasped Time after the onset of the Accident Nuclide 0 hr 1 hr 1 day 1 week 1 month Se-84 7.56E+05 2.54E+00 0.00E+00 0.00E+00 0.00E+00 Se-85 3.77E+05 6.15E-23 0.00E+00 0.00E+00 0.00E+00 Se-87 5.60E+05 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Br-84 6.19E+06 1.67E+06 1.45E-07 2.35E-89 0.00E+00 Br-85 7.14E+06 3.57E+00 0.00E+00 0.00E+00 0.00E+00 Br-87 1.17E+07 4.43E-13 0.00E+00 0.00E+00 0.00E+00 Br-88 1.18E+07 3.86E-60 0.00E+00 0.00E+00 0.00E+00 Br-89 7.95E+06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr-85 6.62E+05 6.62E+05 6.62E+05 6.61E+05 6.59E+05 Kr-85m 1.78E+07 1.53E+07 4.36E+05 9.20E-05 7.46E-42 Kr-87 3.65E+07 2.12E+07 7.60E+01 6.21E-33 0.00E+00 Kr-88 5.16E+07 4.04E+07 1.47E+05 8.04E-11 2.48E-69 Kr-89 6.63E+07 1.28E+02 0.00E+00 0.00E+00 0.00E+00 Kr-90 7.21E+07 2.13E-26 0.00E+00 0.00E+00 0.00E+00 Kr-91 4.97E+07 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Rb-86 4.37E+04 4.36E+04 4.21E+04 3.37E+04 1.43E+04 Rb-87 0.00E+00 4.70E-08 1.12E-07 1.12E-07 1.12E-07 Rb-88 1.58E+07 4.11E+07 1.65E+05 8.98E-11 2.77E-69 Rb-89 2.10E+07 2.57E+06 3.20E-21 0.00E+00 0.00E+00 Rb-90 1.98E+07 9.08E+00 1.12E-96 0.00E+00 0.00E+00 Rb-90m 0.00E+00 7.75E+01 1.90E-95 0.00E+00 0.00E+00 Rb-91 2.49E+07 5.97E-12 0.00E+00 0.00E+00 0.00E+00 Sr-89 1.29E+06 1.30E+06 1.28E+06 1.18E+06 8.60E+05 Sr-90 1.16E+05 1.16E+05 1.16E+05 1.16E+05 1.16E+05 Sr-91 1.73E+06 1.61E+06 3.00E+05 8.21E+00 2.65E-17 Sr-92 1.75E+06 1.36E+06 3.78E+03 3.81E-13 1.84E-74 Sr-95 1.63E+06 3.40E-36 0.00E+00 0.00E+00 0.00E+00 Y-90 1.22E+03 2.45E+03 2.75E+04 9.73E+04 1.16E+05 Y-91 1.58E+04 1.63E+04 2.51E+04 2.54E+04 1.93E+04 Y-91m 1.00E+04 5.44E+05 1.89E+05 5.17E+00 1.67E-17 Y-92 1.77E+04 2.88E+05 3.98E+04 2.97E-08 3.40E-55 12.2-74 Rev. 2

APR1400 DCD TIER 2 Table 12.2-24 (8 of 14)

c. Radioactive Concentration of Post-Accident Airborne in Containment (Bq/cc) (2 of 6)

Elasped Time after the onset of the Accident Nuclide 0 hr 1 hr 1 day 1 week 1 month Y-93 1.27E+04 1.19E+04 2.45E+03 1.25E-01 4.41E-18 Y-95 1.95E+04 3.71E+02 1.01E-37 0.00E+00 0.00E+00 Zr-93 0.00E+00 6.34E-07 7.72E-06 9.56E-06 9.56E-06 Zr-95 1.85E+04 1.85E+04 1.83E+04 1.71E+04 1.34E+04 Zr-97 1.75E+04 1.68E+04 6.54E+03 1.78E+01 2.62E-09 Nb-93m 0.00E+00 1.74E-12 6.34E-10 7.94E-09 3.65E-08 Nb-95 1.84E+04 1.84E+04 1.84E+04 1.83E+04 1.71E+04 Nb-95m 0.00E+00 1.18E+00 2.57E+01 1.05E+02 1.13E+02 Nb-97 0.00E+00 7.38E+03 7.05E+03 1.92E+01 2.82E-09 Nb-97m 0.00E+00 1.59E+04 6.20E+03 1.69E+01 2.48E-09 Nb-99 1.14E+04 1.88E-72 0.00E+00 0.00E+00 0.00E+00 Mo-99 2.41E+05 2.38E+05 1.87E+05 4.13E+04 1.26E+02 Mo-103 2.15E+05 1.86E-13 0.00E+00 0.00E+00 0.00E+00 Tc-99 0.00E+00 8.89E-05 2.00E-03 7.61E-03 9.20E-03 Tc-99m 2.12E+05 2.12E+05 1.81E+05 4.03E+04 1.22E+02 Tc-103 2.17E+05 4.60E-17 0.00E+00 0.00E+00 0.00E+00 Tc-106 1.24E+05 6.37E-25 0.00E+00 0.00E+00 0.00E+00 Ru-103 2.19E+05 2.19E+05 2.15E+05 1.94E+05 1.29E+05 Ru-105 1.66E+05 1.42E+05 3.92E+03 6.76E-07 2.54E-44 Ru-106 9.82E+04 9.82E+04 9.80E+04 9.69E+04 9.28E+04 Rh-103m 0.00E+00 1.14E+05 2.15E+05 1.93E+05 1.29E+05 Rh-105 1.49E+05 1.49E+05 1.07E+05 6.42E+03 1.28E-01 Rh-105m 0.00E+00 3.49E+04 9.62E+02 1.66E-07 6.23E-45 Rh-106 0.00E+00 9.82E+04 9.80E+04 9.69E+04 9.28E+04 Sb-127 2.44E+05 2.42E+05 2.04E+05 6.92E+04 1.10E+03 Sb-129 8.35E+05 7.13E+05 1.90E+04 2.68E-06 4.59E-44 Sb-131 2.05E+06 3.36E+05 2.92E-13 0.00E+00 0.00E+00 Sb-132 1.19E+06 4.22E-01 0.00E+00 0.00E+00 0.00E+00 Sb-133 1.89E+06 5.63E-02 0.00E+00 0.00E+00 0.00E+00 Sb-134 3.64E+05 1.10E-93 0.00E+00 0.00E+00 0.00E+00 Te-127 2.42E+05 2.42E+05 2.24E+05 1.03E+05 3.54E+04 12.2-75 Rev. 2

APR1400 DCD TIER 2 Table 12.2-24 (9 of 14)

c. Radioactive Concentration of Post-Accident Airborne in Containment (Bq/cc) (3 of 6)

Elasped Time after the onset of the Accident Nuclide 0 hr 1 hr 1 day 1 week 1 month Te-127m 4.07E+04 4.07E+04 4.07E+04 3.99E+04 3.49E+04 Te-129 7.96E+05 7.72E+05 1.23E+05 8.93E+04 5.56E+04 Te-129m 1.63E+05 1.63E+05 1.60E+05 1.42E+05 8.82E+04 Te-131 2.08E+06 4.88E+05 6.88E+04 2.47E+03 7.14E-03 Te-131m 5.32E+05 5.20E+05 3.06E+05 1.10E+04 3.17E-02 Te-132 3.55E+06 3.52E+06 2.87E+06 8.01E+05 6.01E+03 Te-133 3.01E+06 2.90E+05 6.30E-03 7.11E-50 0.00E+00 Te-133m 2.51E+06 1.18E+06 3.76E-02 4.24E-49 0.00E+00 Te-134 5.30E+06 1.96E+06 2.26E-04 1.35E-66 0.00E+00 Te-135 2.61E+06 9.41E-51 0.00E+00 0.00E+00 0.00E+00 I-129 1.02E+00 1.02E+00 1.02E+00 1.02E+00 1.02E+00 I-131 2.00E+07 1.99E+07 1.84E+07 1.10E+07 1.51E+06 I-132 2.90E+07 2.24E+07 2.97E+06 8.25E+05 6.19E+03 I-133 4.23E+07 4.10E+07 1.91E+07 1.57E+05 1.61E-03 I-134 4.83+07 2.36E+07 3.94E-01 1.41E-50 0.00E+00 I-135 3.99E+07 3.59E+07 3.22E+06 8.91E-01 6.47E-26 I-137 2.06E+07 1.82E-37 0.00E+00 0.00E+00 0.00E+00 I-138 1.03E+07 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Xe-131m 7.36E+05 7.35E+05 7.06E+05 5.42E+05 1.72E+05 Xe-133 1.05E+08 1.05E+08 9.58E+07 4.56E+07 2.20E+06 Xe-133m 3.23E+06 3.20E+06 2.58E+06 4.37E+05 3.03E+02 Xe-135 3.16E+07 3.25E+07 1.37E+07 3.86E+02 2.23E-16 Xe-135m 2.19E+07 7.17E+06 5.53E+05 1.53E-01 1.11E-26 Xe-137 9.66E+07 1.86E+03 0.00E+00 0.00E+00 0.00E+00 Xe-138 9.77E+07 5.15E+06 2.05E-23 0.00E+00 0.00E+00 Xe-140 5.76E+07 1.19E-72 0.00E+00 0.00E+00 0.00E+00 Cs-134 4.67E+06 4.67E+06 4.67E+06 4.64E+06 4.54E+06 Cs-135 1.01E+01 1.01E+01 1.01E+01 1.01E+01 1.01E+01 Cs-136 1.20E+06 1.20E+06 1.14E+06 8.30E+05 2.47E+05 Cs-137 2.59E+06 2.59E+06 2.59E+06 2.59E+06 2.59E+06 Cs-138 3.09E+07 2.55E+07 3.70E-06 6.24E-87 0.00E+00 12.2-76 Rev. 2

APR1400 DCD TIER 2 Table 12.2-24 (10 of 14)

c. Radioactive Concentration of Post-Accident Airborne in Containment (Bq/cc) (4 of 6)

Elasped Time after the onset of the Accident Nuclide 0 hr 1 hr 1 day 1 week 1 month Cs-140 2.67E+07 2.76E-10 0.00E+00 0.00E+00 0.00E+00 Ba-137m 1.65E+05 2.45E+06 2.45E+06 2.45E+06 2.45E+06 Ba-139 1.98E+06 1.20E+06 1.20E+01 6.05E-31 0.00E+00 Ba-140 1.96E+06 1.96E+06 1.86E+06 1.34E+06 3.86E+05 Ba-143 1.60E+06 3.31E-74 0.00E+00 0.00E+00 0.00E+00 Ba-144 1.33E+06 4.03E-93 0.00E+00 0.00E+00 0.00E+00 La-140 1.96E+04 5.27E+04 6.58E+05 1.42E+06 4.44E+05 La-141 1.81E+04 1.52E+04 2.65E+02 2.64E-09 1.77E-51 La-142 1.81E+04 1.17E+04 5.17E-01 2.82E-28 0.00E+00 La-143 1.79E+04 9.18E+02 1.95E-27 0.00E+00 0.00E+00 La-144 1.63E+04 1.32E-23 0.00E+00 0.00E+00 0.00E+00 Ce-141 4.41E+04 4.41E+04 4.33E+04 3.81E+04 2.33E+04 Ce-143 4.50E+04 4.41E+04 2.72E+04 1.32E+03 1.22E-02 Ce-144 3.20E+04 3.20E+04 3.19E+04 3.15E+04 2.97E+04 Pr-143 1.76E+04 1.77E+04 1.85E+04 1.57E+04 4.89E+03 Pr-144 1.28E+04 3.02E+04 3.19E+04 3.15E+04 2.97E+04 Pr-144m 0.00E+00 4.56E+02 4.56E+02 4.50E+02 4.25E+02 Nd-147 7.07E+03 7.05E+03 6.64E+03 4.54E+03 1.06E+03 Pm-147 0.00E+00 2.13E-01 4.95E+00 2.89E+01 6.79E+01 Sm-147 0.00E+00 7.94E-17 4.48E-14 1.94E-12 2.38E-11 Tl-207 0.00E+00 3.42E-23 4.46E-23 1.44E-23 2.44E-23 Tl-208 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6.29E-23 Tl-209 0.00E+00 0.00E+00 1.01E-21 8.81E-22 1.85E-22 Pb-209 0.00E+00 2.40E-22 2.13E-22 1.88E-20 1.33E-20 Pb-210 0.00E+00 0.00E+00 1.10E-17 1.27E-17 0.00E+00 Pb-211 0.00E+00 0.00E+00 3.88E-24 0.00E+00 4.44E-24 Pb-212 0.00E+00 8.59E-24 0.00E+00 0.00E+00 2.70E-22 Pb-214 0.00E+00 2.99E-17 3.00E-17 4.98E-17 1.28E-16 Bi-210 0.00E+00 2.15E-18 1.26E-17 1.88E-17 1.35E-17 Bi-211 0.00E+00 4.85E-23 4.67E-22 0.00E+00 3.47E-23 Bi-213 0.00E+00 7.55E-24 2.85E-20 1.26E-19 6.95E-20 12.2-77 Rev. 2

APR1400 DCD TIER 2 Table 12.2-24 (11 of 14)

c. Radioactive Concentration of Post-Accident Airborne in Containment (Bq/cc) (5 of 6)

Elasped Time after the onset of the Accident Nuclide 0 hr 1 hr 1 day 1 week 1 month Bi-214 0.00E+00 3.15E-17 3.34E-17 4.25E-17 1.29E-16 Po-210 0.00E+00 1.64E-17 2.54E-17 3.76E-17 2.86E-17 Po-211 0.00E+00 4.18E-26 6.62E-26 0.00E+00 4.43E-27 Po-213 0.00E+00 7.39E-24 2.79E-20 1.23E-19 6.80E-20 Po-214 0.00E+00 1.06E-17 1.20E-17 2.11E-17 1.07E-16 Po-215 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.43E-23 Po-216 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.46E-22 Po-218 0.00E+00 1.74E-17 2.75E-17 3.53E-17 1.28E-16 At-217 0.00E+00 5.92E-20 7.92E-20 1.20E-20 3.64E-21 Rn-220 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.48E-22 Rn-222 0.00E+00 1.65E-17 6.97E-18 1.15E-17 1.10E-16 Fr-221 0.00E+00 0.00E+00 3.38E-19 1.13E-19 1.95E-20 Fr-223 0.00E+00 0.00E+00 0.00E+00 0.00E+00 9.75E-24 Ra-224 0.00E+00 1.29E-23 0.00E+00 1.64E-23 3.66E-23 Ra-225 0.00E+00 1.06E-19 6.65E-20 0.00E+00 6.79E-20 Ra-226 0.00E+00 9.61E-18 2.04E-17 3.72E-17 1.86E-16 Ra-228 0.00E+00 1.96E-23 0.00E+00 0.00E+00 1.89E-21 Ac-225 0.00E+00 5.06E-20 2.99E-22 9.45E-21 5.38E-20 Ac-227 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.04E-21 Ac-228 0.00E+00 9.75E-22 0.00E+00 6.08E-24 1.11E-21 Th-228 0.00E+00 2.82E-22 0.00E+00 1.31E-21 1.74E-21 Th-229 0.00E+00 6.64E-20 5.03E-20 1.80E-19 1.35E-19 Th-230 0.00E+00 2.51E-17 1.57E-14 7.69E-13 1.41E-11 Th-231 0.00E+00 1.19E-14 5.66E-12 1.18E-10 6.17E-10 Th-232 0.00E+00 3.08E-23 0.00E+00 3.88E-21 7.37E-20 Pa-231 0.00E+00 0.00E+00 0.00E+00 1.96E-17 5.09E-16 Pa-233 0.00E+00 3.56E-14 2.04E-11 9.80E-10 1.65E-08 U-233 0.00E+00 0.00E+00 1.70E-19 3.37E-17 2.03E-15 U-234 0.00E+00 5.30E-08 1.27E-06 8.91E-06 3.82E-05 U-235 0.00E+00 8.88E-13 2.14E-11 1.51E-10 6.50E-10 U-236 0.00E+00 5.00E-11 1.20E-09 8.40E-09 3.60E-08 12.2-78 Rev. 2

APR1400 DCD TIER 2 Table 12.2-24 (12 of 14)

c. Radioactive Concentration of Post-Accident Airborne in Containment (Bq/cc) (6 of 6)

Elasped Time after the onset of the Accident Nuclide 0 hr 1 hr 1 day 1 week 1 month U-237 0.00E+00 4.37E-04 9.99E-03 5.25E-02 9.74E-02 Np-237 0.00E+00 6.65E-11 1.61E-09 1.18E-08 5.71E-08 Np-239 5.81E+05 5.74E+05 4.33E+05 7.40E+04 8.50E+01 Pu-238 1.64E+02 1.64E+02 1.64E+02 1.64E+02 1.64E+02 Pu-239 7.90E+00 7.90E+00 7.94E+00 8.04E+00 8.06E+00 Pu-240 1.48E+01 1.48E+01 1.48E+01 1.48E+01 1.48E+01 Pu-241 4.18E+03 4.18E+03 4.18E+03 4.18E+03 4.16E+03 Am-241 1.80E+00 1.80E+00 1.82E+00 1.93E+00 2.35E+00 Cm-242 9.01E+02 9.01E+02 8.97E+02 8.75E+02 7.93E+02 Cm-244 1.28E+02 1.28E+02 1.28E+02 1.28E+02 1.28E+02 12.2-79 Rev. 2

APR1400 DCD TIER 2 Table 12.2-24 (13 of 14)

d. MCR Emergency Makeup ACU Filter Inventories (Bq) (1 of 2)

Elasped Time After the Onset of the Accident Nuclide 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1 day 4 days 1 week 1 month I-131 2.77E+09 1.75E+10 2.25E+10 2.48E+10 3.12E+10 I-132 3.70E+09 2.33E+10 2.99E+10 3.31E+10 4.16E+10 I-133 5.30E+09 3.34E+10 4.29E+10 4.75E+10 5.96E+10 I-134 5.19E+08 3.28E+09 4.22E+09 4.67E+09 5.87E+09 I-135 3.96E+09 2.49E+10 3.20E+10 3.55E+10 4.45E+10 Co-58 1.67E+05 9.05E+05 9.54E+05 9.72E+05 9.90E+05 Co-60 1.28E+05 6.93E+05 7.30E+05 7.44E+05 7.58E+05 Rb-86 5.68E+06 3.08E+07 3.24E+07 3.31E+07 3.37E+07 Sr-89 1.99E+08 1.08E+09 1.14E+09 1.16E+09 1.18E+09 Sr-90 1.79E+07 9.69E+07 1.02E+08 1.04E+08 1.06E+08 Sr-91 2.11E+08 1.14E+09 1.20E+09 1.23E+09 1.25E+09 Sr-92 1.17E+08 6.32E+08 6.67E+08 6.79E+08 6.92E+08 Y-90 2.46E+05 1.33E+06 1.40E+06 1.43E+06 1.46E+06 Y-91 2.46E+06 1.33E+07 1.40E+07 1.43E+07 1.46E+07 Y-92 9.73E+06 5.27E+07 5.56E+07 5.67E+07 5.77E+07 Y-93 1.56E+06 8.46E+06 8.91E+06 9.09E+06 9.25E+06 Zr-95 2.85E+06 1.54E+07 1.63E+07 1.66E+07 1.69E+07 Zr-97 2.37E+06 1.28E+07 1.35E+07 1.38E+07 1.40E+07 Nb-95 2.84E+06 1.54E+07 1.62E+07 1.66E+07 1.69E+07 Mo-99 3.59E+07 1.95E+08 2.05E+08 2.09E+08 2.13E+08 Tc-99m 3.26E+07 1.77E+08 1.86E+08 1.90E+08 1.93E+08 Ru-103 3.38E+07 1.83E+08 1.93E+08 1.97E+08 2.00E+08 Ru-105 1.54E+07 8.33E+07 8.78E+07 8.95E+07 9.11E+07 Ru-106 1.51E+07 8.21E+07 8.65E+07 8.82E+07 8.97E+07 Rh-105 2.28E+07 1.24E+08 1.30E+08 1.33E+08 1.35E+08 Sb-127 3.67E+07 1.99E+08 2.10E+08 2.14E+08 2.18E+08 Sb-129 7.61E+07 4.12E+08 4.34E+08 4.43E+08 4.51E+08 Te-127 3.72E+07 2.02E+08 2.13E+08 2.17E+08 2.21E+08 Te-127m 6.30E+06 3.41E+07 3.60E+07 3.67E+07 3.73E+07 Te-129 9.43E+07 5.11E+08 5.39E+08 5.49E+08 5.59E+08 Te-129m 2.52E+07 1.36E+08 1.44E+08 1.47E+08 1.49E+08 12.2-80 Rev. 2

APR1400 DCD TIER 2 Table 12.2-24 (14 of 14)

d. MCR Emergency Makeup ACU Filter Inventories (Bq) (2 of 2)

Elasped Time After the Onset of the Accident Nuclide 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1 day 4 days 1 week 1 month Te-131m 7.62E+07 4.13E+08 4.35E+08 4.44E+08 4.52E+08 Te-132 5.32E+08 2.88E+09 3.04E+09 3.10E+09 3.16E+09 Cs-134 6.11E+08 3.31E+09 3.49E+09 3.56E+09 3.62E+09 Cs-136 1.56E+08 8.45E+08 8.90E+08 9.07E+08 9.24E+08 Cs-137 3.39E+08 1.83E+09 1.93E+09 1.97E+09 2.01E+09 Ba-139 5.88E+07 3.19E+08 3.36E+08 3.42E+08 3.49E+08 Ba-140 3.00E+08 1.63E+09 1.72E+09 1.75E+09 1.78E+09 La-140 4.57E+06 2.47E+07 2.61E+07 2.66E+07 2.71E+07 La-141 1.56E+06 8.46E+06 8.92E+06 9.09E+06 9.25E+06 La-142 6.40E+05 3.47E+06 3.65E+06 3.72E+06 3.79E+06 Ce-141 6.82E+06 3.69E+07 3.89E+07 3.97E+07 4.04E+07 Ce-143 6.50E+06 3.52E+07 3.71E+07 3.78E+07 3.85E+07 Ce-144 4.94E+06 2.68E+07 2.82E+07 2.88E+07 2.93E+07 Pr-143 2.73E+06 1.48E+07 1.56E+07 1.59E+07 1.61E+07 Nd-147 1.08E+06 5.86E+06 6.18E+06 6.30E+06 6.41E+06 Np-239 8.63E+07 4.67E+08 4.93E+08 5.02E+08 5.11E+08 Pu-238 2.54E+04 1.38E+05 1.45E+05 1.48E+05 1.51E+05 Pu-239 1.22E+03 6.62E+03 6.98E+03 7.11E+03 7.24E+03 Pu-240 2.30E+03 1.24E+04 1.31E+04 1.34E+04 1.36E+04 Pu-241 6.45E+05 3.50E+06 3.69E+06 3.76E+06 3.82E+06 Am-241 2.78E+02 1.50E+03 1.59E+03 1.62E+03 1.65E+03 Cm-242 1.39E+05 7.53E+05 7.94E+05 8.09E+05 8.23E+05 Cm-244 1.98E+04 1.07E+05 1.13E+05 1.15E+05 1.18E+05 12.2-81 Rev. 2

APR1400 DCD TIER 2 Table 12.2-25 (1 of 5)

Radioactive Source Dimensions and Parameters Used in Shielding Analysis Source Dimension Source Characteristic Housing Diameter Length Partial Density Thickness Building Component Shape (or Width) (cm) (cm) Height (cm) Material (g/cm3) Material (cm)

Reactor Pressurizer Cylinder 244.48 - 700.06 Water: 48% 0.59 Steel 12.38 Containment 747.71 Vapor: 52% 0.001293 Building Reactor coolant pump Cylinder 185.00 - 126.74 Water: 100 % 0.75 Steel 14.00 Reactor drain tank Cylinder Liquid: 162.90 - 528.57 Water: 73 % 1.00 Not considered Vapor: 99.70 Vapor: 27 % 0.001293 Regenerative HX Cylinder 24.69 - 400.69 Water: 85 % 0.85 Steel 2.22 Steel: 15 % 1.18 Letdown HX Cylinder 45.72 - 341.36 Water: 88 % 0.88 Steel 2.54 Steel: 12 % 0.95 Steam generator Annular OD: 497.80 - 969.57 Water: 100 % 0.70 Steel 12.86 cylinder ID: 415.80 Semisphere 472.60 Auxiliary SC HX Cylinder 137.16 - 803.15 Water: 94 % 0.94 Steel 1.27 Building Steel: 6 % 0.54 SC miniflow HX(1) Cylinder 33.66 - 173.43 Water: 93 % 0.93 Steel 0.95 Steel: 7 % 0.59 SC pump Cylinder 38.1 - 609.6 Water: 100% 1 Steel 1.27 Charging pump Cylinder 38.10 - 298.70 Water: 94 % 0.94 Steel 1.27 miniflow HX Steel: 6 % 0.50 CS HX(1) Cylinder 129.54 - 701.04 Water: 94 % 0.94 Steel 1.59 Steel: 6 % 0.49 (1) For post-accident, the volume to estimate source term in the heat exchanger is limited to the volume of tube side. (SC mini flow HX : 4.33E+04 cm3, CS HX : 3.27E+06 cm3) 12.2-82 Rev. 2

APR1400 DCD TIER 2 Table 12.2-25 (2 of 5)

Source Dimension Source Characteristic Housing Diameter Length Height Partial Density Thickness Building Component Shape (or Width) (cm) (cm) (cm) Material (g/cm3) Material (cm)

Auxiliary CS miniflow HX(1) Cylinder 31.75 - 186.06 Water: 94 % 0.94 Steel 0.95 Building Steel: 6 % 0.45 CS pump Cylinder 38.1 - 609.6 Water: 100 % 1 Steel 1.27 SI pump Cylinder 25.5 - 609.6 Water: 100 % 1 Steel 0.93 Equipment drain tank Cylinder Liquid: 193.59 - 610.87 Water: 50 % 1.00 Not considered (Horizontal) Vapor: 193.59 (Length of tank) Vapor: 50 % 0.001293 Boric acid Cylinder Liquid: 193.53 - 180.52 Water: 47 % 1.00 Not considered concentrator Vapor: 206.58 Vapor: 53 % 0.001293 SFP cleanup demin. Cylinder 145.70 - 144.17 Water: 100 % 1.00 Not considered Boric acid condensate Cylinder 74.60 - 206.17 Water: 100 % 1.00 Not considered IX Deborating IX Cylinder 105.08 - 104.49 Water: 100 % 1.00 Not considered Pre-holdup IX Cylinder 105.08 - 104. 49 Water: 100 % 1. 00 Not considered Purification IX Cylinder 105.08 - 104.49 Water: 100 % 1.00 Not considered SFP cooling HX Rectangular 31.15 134.15 198.28 Water: 67 % 0.67 Not considered Steel: 33 % 2.63 Volume control tank Cylinder 241.44 Liquid: 223.23 Water: 40 % 1.00 Not considered Vapor: 330.72 Vapor: 60 % 0.001293 (1) For post-accident, the volume to estimate source term in the CS mini flow heat exchanger is limited to the volume of tube side, 3.68E+04 cm3.

12.2-83 Rev. 2

APR1400 DCD TIER 2 Table 12.2-25 (3 of 5)

Source Dimension Source Characteristic Housing Diameter Length Height Partial Density Thickness Building Component Shape (or Width) (cm) (cm) (cm) Material (g/cm3) Material (cm)

Auxiliary SGBD flash tank Cylinder 304.8 - 455.96 Water: 100 % 1.00 Not considered Building SGBD HX Cylinder 84.86 - 487.68 Water: 86 % 0.90 Steel SGBD Steel: 14 % 1.12 HX Spent fuel pool Rectangular 869.00 1,113.50 381.00 Water: 70% 0.70 Not considered UO2: 22% 1.98 Zircaloy: 8%

0.56 Cask loading pit Rectangular 20.23 20.23 381.00 Water: 58% 0.58 Not considered parallelepiped UO2: 30% 2.76 Zircaloy: 12% 0.79 GRS header drain tank Cylinder 45.72 - 172.48 Vapor: 100% 0.001293 Not considered IRWST Annular O.D: 4378.96 - 358.04 Water: 77% 1.00 Not considered Cylinder I.D: 3230.88 122.24 Vapor: 23% 0.001293 Seal injection filter Cylinder 10.80 - 54.00 Water: 100% 1.00 Not considered Reactor drain filter Cylinder 19.35 - 50.50 Water: 100% 1.00 Not considered Boric acid filter Cylinder 19.35 - 50.50 Water: 100% 1.00 Not considered Purification filter Cylinder 19.35 - 50.50 Water: 100% 1.00 Not considered Reactor makeup water Cylinder 19.35 - 50.50 Water: 100% 1.00 Not considered filter Steam generator Cylinder 228.60 - 152.40 Water: 100% 1.00 Not considered blowdown mixed-bed 12.2-84 Rev. 2

APR1400 DCD TIER 2 Table 12.2-25 (4 of 5)

Source Dimension Source Characteristic Housing Diameter Length Height Partial Density Thickness Building Component Shape (or Width) (cm) (cm) (cm) Material (g/cm3) Material (cm)

Auxiliary Blowdown pre-filter Cylinder 25.93 - 51.12 Water: 100% 1.00 Not considered Building Blowdown post-filter Cylinder 25.93 - 51.12 Water: 100% 1.00 Not considered Compound Chemical waste tank Cylinder 304.80 - 466.91 Water: 100% 1.00 Not considered Building Floor drain tank Cylinder 358.14 - 676.38 Water: 100% 1.00 Not considered Low-activity spent Cylinder 274.32 - 383.33 Water: 100% 1.00 Not considered resin tank Spent resin long-term Cylinder 487.68 - 482.92 Water: 100% 1.00 Not considered storage tank Waste drum storage Rectangular 601.98 782.57 262.89 Carbon: 2.62 Not considered parallelepiped 100%

Header Drain Tank Cylinder 76.20 - 121.92 Vapor: 100% 0.001293 steel 0.6 Waste Gas Dryer Cylinder 30.318 - 91.72 Vapor: Steel 1.03 0.001009 82.73%

1.357 Steel: 17.27%

Guard Bed Cylinder 50.80 - 139.71 Carbon: 100% 0.41 Not considered Delay Bed Cylinder 180.08 - 456.05 Carbon: 100% 0.41 Not considered 12.2-85 Rev. 2

APR1400 DCD TIER 2 Table 12.2-25 (5 of 5)

Source Dimension Source Characteristic Housing Diameter Length Height Partial Density Thickness Building Component Shape (or Width) (cm) (cm) (cm) Material (g/cm3) Material (cm)

Compound HEPA filter Cylinder 45.72 - 50.80 Vapor: 100% 0.001293 Not considered Building Equipment waste tank Cylinder 358.14 - 676.38 Water: 100% 1.00 Not considered Monitor tank Cylinder 487.68 - 547.16 Water: 100% 1.00 Not considered Reverse osmosis Cylinder 66.33 - 101.60 Water: 100% 1.00 Not considered LRS IX (Cation Bed) Cylinder 120.17 - 124.83 Water: 100% 1.00 Not considered LRS IX (Mixed1 Cylinder 120.17 124.83 Water: 100% 1.00 Not considered Bed)

LRS IX (Mixed2 Cylinder 120.17 124.83 Water: 100% 1.00 Not considered Bed)

Turbine CPS cation bed Sphere 314.96 - - Water: 100% 1.00 Steel 3.40 Building CPS mixed bed Sphere 314.96 - - Water: 100% 1.00 Steel 3.40 Yard Boric Acid Storage Cylinder 1342.12 - Liquid : 334.96 Water: 50% 1.00 Concrete 40.64 Tank Vapor : 334.96 Vapor: 50% 0.001293 Holdup Tank Cylinder 1706.88 - Liquid : 86.85 Water: 12.5% 1.00 Steel 0.635 Vapor : 607.96 Vapor: 87.5% 0.001293 Concrete 37.465 Reactor Makeup Cylinder 1676.40 - Liquid : 521.62 Water: 77% 1.00 Not considered Water Tank Vapor : 155.81 Vapor: 23% 0.001293 12.2-86 Rev. 2

APR1400 DCD TIER 2 Table 12.2-26 (1 of 8)

Assumptions and Parameters Used in Airborne Source Term Calculations Reactor Containment Building Assumptions/Parameters Values Fuel failure 0.25 % fuel defect RCS leak rate 1.89 L/min (0.5 gpm)

(design basis leak rate)

RCS letdown flow 302.8 L/min (80 gpm)

Iodine spike Not considered Low-volume purge flow rate 2.549E + 03 m3/hr (1,500 cfm)

(Assumed that CLVPS operates for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> before shutdown)

High-volume purge flow rate 9.14E + 04 m3/hr (54,000 cfm)

(Assumes that CHVPS operates at cold shutdown operation mode for 29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br />)

Filter efficiency for low-volume Particulate: 99 %

purge filters Halogen: 99 %

Gas stripping No gas stripping Containment free volume 8.858E + 04 m3 (3.128E + 06 ft3) 12.2-87 Rev. 2

APR1400 DCD TIER 2 Table 12.2-26 (2 of 8)

Fuel Handling Area in Auxiliary Building Assumptions/Parameters Values 2

Surface Area of SFP 1.3852E+02 m SFP Water Volume 1.72E+03 m3 (=649,000 gal)

SFP Temperature 49 ºC (120 °F) (during power operation) 60 ºC (140 °F) (during refueling and shutdown)

Fuel Handling Area HVAC flow (1) 28,450 cfm Air Pressure Above SFP 1.033 kg/cm2 Air Temperature Above SFP 25 ºC (77 °F)

Relative Humidity Above SFP 70 %

Wind Speed on the Water Surface 0.101 m/s (20 ft/min)

RCS Water Mass 3.0E+05 kg Tritium Production Rate in 9.92E+13 Bq/1.5 yr (=6.613E+13 Bq/yr)

Reactor Coolant RCS Water Density 700 kg/m3 Evaporation Rate of SFP 1.697E-02 m3/hr (1) The SFP area includes the spent fuel pool, cask loading pits, fuel loading & unloading, refueling canal, and new fuel container laydown & inspection area 12.2-88 Rev. 2

APR1400 DCD TIER 2 Table 12.2-26 (3 of 8)

Auxiliary and Compound Building Volume Leak Sources and Leak Rate Source Minimum Required Ventilation Cubicle (m3) Number of Sources (L/min) Terms (1), (2) Flow (m3/hr)

CS Pump and Miniflow 461 Pump seal (1) 8.33E-04 IRWST 170 HX Rm Flange (4) 2.00E-03 (050-A01C, D) Valve 1 (4) 6.66E-04 Valve 4 (2) 1.33E-03 Valve 14 (1) 2.33E-03 SI Pump Rm 610 Pump seal (1) 1.67E-02 IRWST 340 (050-A02C, D) Flange (3) 1.50E-03 Valve 4 (4) 2.66E-03 Valve 10 (1) 1.67E-03 Floor Drain Sump 94.9 Evaporation from 6.74E-02 0.1 PCA 680 Pump Rm sump (055-A34A, B, C, D)

Pipe Chase and Valve Rm 996 Valve 18 (2) 6.00E-03 IRWST 170 (055-A14C) Valve 20 (1) 3.33E-03 Shutdown Cooling HX Rm 925 Valve 1.5 (1) 2.50E-04 IRWST 170 (055-A30A, B) Valve 10 (2) 3.33E-03 12.2-89 Rev. 2

APR1400 DCD TIER 2 Table 12.2-26 (4 of 8)

Volume Leak Sources and Leak Rate Source Terms Minimum Required Ventilation Cubicle (m3) Number of Sources (L/min) (1), (2)

Flow (m3/hr)

Charging Pump Rm 833 Pump Seal (1) 8.33E-04 VCT - 850 (055-A55B) Flange (6) 3.00E-03 Halogens Valve 1 (1) 1.67E-04 RCS - NG Valve 3 (4) 2.00E-03 Valve 4 (2) 1.33E-03 Charging Pump 187 Valve 2.5 (1) 4.16E-04 VCT - 170 Miniflow HX Rm Halogens (055-A43A) RCS - NG Equipment Drain Tank Rm 501 Flange (2) 9.99E-04 EDT 170 (055-A51B) Valve 1 (1) 1.67E-04 Valve 6 (1) 9.99E-04 Reactor Drain Pump Rm 200 Pump Seal (1) 8.33E-04 RDT 850 (055-A52B, A53B) Flange (4) 2.00E-03 Gas Stripper Rm 697 Flange (1) 5.00E-04 Gas Stripper 1,444 (068-A06A)

Filter and Demi. Valve Area 174 Valve 3 (1) 5.00E-04 RDT 1,529 (068-A10A) Valve 3 (27) 1.35E-02 1.0 PCA 12.2-90 Rev. 2

APR1400 DCD TIER 2 Table 12.2-26 (5 of 8)

Volume Leak Sources and Leak Rate Source Minimum Required Ventilation Cubicle (m3) Number of Sources (L/min) Terms (1), (2) Flow (m3/hr)

SFP Cleanup Pump Rm 224 Flange (6) 5.00E-03 SFP 510 (078-A38A) Valve 4 (3) 2.00E-03 Valve 6 (4) 4.01E-03 Reactor Makeup Water 168 Flange (8) 4.00E-03 RMWT 340 Pump Rm (078-A49B) Pump Seal (2) 1.67E-03 Holdup Pump Rm 311 Flange (8) 4.00E-03 Holdup Tank 340 (078-A50B) Pump Seal (2) 1.67E-03 Valve Rm (120-A23A) 77 Valve 3 (1) 5.00E-04 1.0 PCA 510 Valve 3 (1) 5.00E-04 RMWT Valve 3 (1) 5.00E-04 BAST Valve Rm (063-P07) 275 Flange (4) 2.00E-03 1.0 PCA 340 Valve 2 (3) 9.99E-04 Without Valve 2.5 (2) 8.33E-04 Noble Gas Flange (2) 9.99E-04 1.0 PCA Valve 2 (6) 2.00E-03 Equipment Waste Pump Rm 138 Pump Seal (1) 8.33E-04 0.32 PCA 340 (063-P21, P22) Flange (6) 3.00E-03 Valve 2 (2) 6.66E-04 Valve 3 (6) 3.00E-03 Equipment Waste Tank Rm 221 Flange (2) 9.99E-04 Equipment 340 (063-R23, P24) Waste Tank 12.2-91 Rev. 2

APR1400 DCD TIER 2 Table 12.2-26 (6 of 8)

Volume Leak Sources and Leak Rate Source Minimum Required Ventilation Cubicle (m3) Number of Sources (L/min) Terms (1), (2) Flow (m3/hr)

Floor Drain Pump Rm 413 Pump seal (1) 8.33E-04 0.44 PCA 340 (063-P25) Flange (2) 9.99E-04 Valve 2 (4) 1.33E-03 Valve 3 (12) 6.00E-03 Normal Sump Pump Rm 68.5 Flange (4) 2.00E-03 0.01 PCA 510 (063-P26) Valve 3 (2) 9.99E-04 Evaporation from 4.92E-02 sump Chemical Waste Pump Rm 215 Pump seal (2) 1.67E-02 0.01 PCA 1,020 (063-P27) Flange (10) 5.00E-03 Valve 3 (11) 5.49E-03 Valve 2 (2) 6.66E-04 Floor Drain Tank Rm 184 Flange (2) 9.99E-04 Floor Drain 170 (063-P28, P29) Tank Chemical Waste Tank Rm 331 Flange (2) 9.99E-04 Chemical 170 (063-P30, P31) Waste Tank 12.2-92 Rev. 2

APR1400 DCD TIER 2 Table 12.2-26 (7 of 8)

Volume Leak Sources and Leak Rate Source Terms Minimum Required Ventilation Cubicle (m3) Number of Sources (L/min) (1), (2)

Flow (m3/hr)

Detergent Waste Tank and 657 Pump seal(2) 1.67E-02 Detergent 170 Pump Rm Flange (4) 2.00E-03 Waste Tank (063-P32) Valve 2 (2) 6.66E-04 Valve 3 (3) 1.50E-03 Valve 4 (7) 4.66E-03 Evaporation from 3.51E-02 sump Chemical Drain Sump Pump 119 Flange (2) 9.99E-04 0.01 PCA 934 Rm Valve 2 (1) 3.33E-04 (063-P36) Evaporation from 4.92E-02 sump Monitor Tank Rm 885 Flange (4) 2.00E-03 Monitor Tank 170 (063-P37)

Monitor Tank Pump Rm 197 Pump seal (2) 1.67E-02 Monitor Tank 170 (063-P54) Flange (10) 5.00E-03 Valve 2.5 (9) 3.75E-03 Valve 3 (2) 9.99E-04 Valve 4 (14) 9.33E-03 Valve Rm (085-P06) 180 Valve 4 (1) 6.66E-04 1.0 PCA 170 (Except for NG) 12.2-93 Rev. 2

APR1400 DCD TIER 2 Table 12.2-26 (8 of 8)

Volume Leak Sources and Leak Rate Source Terms Minimum Required Ventilation Cubicle (m3) Number of Sources (L/min) (1), (2)

Flow (m3/hr)

Valve Rm (085-P15) 263 Valve 3 (1) 5.00E-04 1 PCA 510 Valve 6 (4) 5.00E-03 (Except for NG)

Valve Rm (085-P16) 269 Valve 4 (1) 6.66E-04 0.32 PCA 1,530 Valve 6 (2) 2.00E-03 0.32 PCA Valve 6 (3) 3.00E-03 0.1 PCA Valve 6 (1) 9.99E-04 0.44 PCA Valve 4 (2) 1.33E-03 0.01 PCA (1) PCA: Fraction of primary coolant activity concentrations (2) NG: Noble gases (3) The HVAC flows listed in this table represent the actual minimum flow rates required for the ventilation for the corresponding rooms 12.2-94 Rev. 2

APR1400 DCD TIER 2 Table 12.2-27 Vapor Activities in Pressurizer Parameters Used in Pressurizer Vapor Activity Calculation Parameter Value Liquid Volume (cm3) 3.32E+07 Vapor Volume (cm3) 3.57E+07 Liquid Mass (g) 1.99E+07 Vapor Mass (g) 3.52E+06 Temperature (°K) 619.44 Henrys Constant (Xe), atm 9.21E+03 Henrys Constant (Kr), atm 1.84E+03 Liquid specific volume, (cm3/g)(1) 1.67 3 (1)

Vapor specific volume, (cm /g) 10.16 Pressurizer Vapor Activities Nuclide Activity (Bq/g)

Kr-85m 7.5E+04 Kr-85 3.2E+05 Kr-87 5.9E+04 Kr-88 1.6E+05 Xe-131m 8.0E+06 Xe-133m 4.9E+05 Xe-133 5.2E+08 Xe-135m 1.1E+06 Xe-135 1.1E+07 Xe-137 2.5E+05 Xe-138 9.3E+05 3

(1) Saturation condition at 154.7 kg/cm A (2,200 psia) 12.2-95 Rev. 2

APR1400 DCD TIER 2 Table 12.2-28 (1 of 2)

Parameters Used in CVCS Inventory Calculations Components Parameters Value Components Parameters Value 3 3 Letdown HX Tube Volume (cm ) 270658 Equipment Tank Volume (cm ) 3.596E+07 Drain Tank Regenerative Tube Volume (cm3) 27984 Low Level Fraction 0.08 HX Shell Volume (cm3) 174886 High Level Fraction 0.37 CCP Mini Tube Volume 145738 RCS Activity 0.1 Flow HX Fraction Purification Vessel Volume (cm3) 1578517 Flow rate (cm3/sec) 2.19 IX Resin Volume (cm3) 906138 Volume Tank Volume (cm3) 2.536E+07 Control Tank Deborating Vessel Volume (cm3) 1578517 Low Level Fraction 0.31 IX Resin Volume (cm3) 906138 High Level Fraction 0.62 Pre-holdup Vessel Volume (cm3) 1578517 Holdup Tank Tank Volume (cm3) 1.590E+09 IX Resin Volume (cm3) 906138 Low Level Fraction 0.05 Boric Acid Vessel Volume (cm3) 1457384 High Level Fraction 0.20 Condensate IX Resin Volume (cm3) 906138 Reactor Tank Volume (cm3) 1.495E+09 Makeup Water Seal Vessel Volume (cm3) 7570 Tank Low Level Fraction 0.58 Injection Filter Reactor Vessel Volume (cm3) 15142 High Level Fraction 0.95 Drain Filter Boric Acid Vessel Volume (cm3) 15142 Boric Acid Tank Volume (cm3) 9.464E+08 Filter Storage Tank Purification Vessel Volume (cm3) 15142 Low Level Fraction 0.40 Filter Reactor Vessel Volume (cm3) 15142 High Level Fraction 0.95 Makeup Water Filter Reactor Tank Volume (cm3) 1.514E+07 In-containment Tank Volume (cm3) 3.295E+09 Drain Tank Refueling Low Level Fraction 0.40 Water Storage Low Level Fraction 0.76 Tank High Level Fraction 0.79 High Level Fraction 0.78 Flow rate (cm3/sec) 11 12.2-96 Rev. 2

APR1400 DCD TIER 2 Table 12.2-28 (2 of 2)

Components Parameters Value Remarks Reactor Drain Tank Partition Factor 0.001 Particulates 10 Noble Gas Equipment Drain Tank Partition Factor 0.0001 Particulates 10 Noble Gas Volume Control Tank Partition Factor 0.001 Particulates 10 Noble Gas Holdup Tank Partition Factor 0.0001 Particulates 10 Noble Gas Reactor Makeup Water Tank Partition Factor 0.0001 Particulates 10 Noble Gas Boric Acid Storage Tank Partition Factor 0.0001 Particulates 10 Noble Gas In-containment Refueling Water Partition Factor 0.0001 Particulates Storage Tank 10 Noble Gas Yearly Generated Waste Volume (cm3) 4.164E+09 Boration Waste Volume (cm3) 1.772E+08 Dilution Waste Volume (cm3) 3.358E+08 Drained Coolant Volume (cm3) 1.616E+08 Contraction Addition Volume (cm3) 1.268E+08 RCS Partial Refill Addition Volume (cm3) 4.823E+08 12.2-97 Rev. 2

APR1400 DCD TIER 2 1.0E+02 1.0E+01 1.0E+00 1.0E-01 Dose Rate [mSv/hr]

1.0E-02 1.0E-03 1.0E-04 72 Hour Decay 1.0E-05 100 Hour Decay 200 Hour Decay 1.0E-06 1000 Hour Decay 2000 Hour Decay 1.0E-07 100 150 200 250 300 350 400 Axial Height from Top of Active Fuel [cm]

Note: Shielding effect of the refueling machine is included.

Figure 12.2-1 Spent Fuel Assembly Dose Rate vs. Axial Distance in Refueling Pool 12.2-98 Rev. 2

APR1400 DCD TIER 2 1.0E+08 1.0E+07 1.0E+06 1.0E+05 1.0E+04 Dose Rate [mSv/hr]

1.0E+03 1.0E+02 1.0E+01 1.0E+00 72 Hour Decay 1.0E-01 100 Hour Decay 200 Hour Decay 1.0E-02 1000 Hour Decay 1.0E-03 2000 Hour Decay 1.0E-04 0 50 100 150 200 250 300 350 400 Radial Distance from Edge of Active Fuel [cm]

Note: Shielding effect of the refueling machine is included.

Figure 12.2-2 Spent Fuel Assembly Dose Rate vs. Radial Distance in Refueling Pool 12.2-99 Rev. 2

APR1400 DCD TIER 2 12.3 Radiation Protection Design Features 12.3.1 Facility Design Features The APR1400 design incorporates as low as (is) reasonably achievable (ALARA) principles per U.S. Nuclear Regulatory Commission (NRC) Regulatory Guides (RGs) 8.8 (Reference 1) and 8.10 (Reference 2) to minimize onsite exposures to plant personnel during normal operation and accident conditions. Subsection 12.1.2 describes the approach for incorporating the ALARA principles into the plant layout, equipment selection and arrangement, and material selection. The plant layout is designed to provide adequate shielding between high-radiation areas and the low-radiation areas that are occupied by plant personnel. In addition, cubicles/rooms with similar radiation zone designations are grouped to the extent possible. These design approaches are implemented to provide reasonable assurance that occupational exposures are maintained below the limits established in 10 CFR Part 20 (Reference 3) and are maintained ALARA.

The following subsections detail the design features that provide reasonable assurance that operational and maintenance exposures are ALARA.

12.3.1.1 General Arrangement Design Features

a. Locations of radioactive systems and equipment Nonradioactive systems are physically separated from radioactive systems to the extent possible. This approach helps control the spread of contamination and minimizes the necessity of routing piping containing radioactive fluids or slurries through nonradioactive areas such as personnel corridors. The approach also facilitates access control of the radiation control areas and is consistent with NRC RG 8.8 Position 2.a and NRC RG 4.21 (Reference 4).

Radioactive components are located in separate compartments wherever possible.

The compartment design takes into consideration the frequency of access, operational requirements, and radiation level. For example, ion exchangers containing radioactive resins are located in separate compartments, and although the valves are remotely operated, they are located in a separate shielded gallery in the event manual operation or access for maintenance activities is needed. The compartment walls provide shielding, which enables personnel to perform operation and maintenance activities in a lower radiation area with significantly 12.3-1 Rev. 2

APR1400 DCD TIER 2 decreased radiation exposure. This approach is consistent with NRC RG 8.8, Position 2.

b. Pipe routing Components with similar levels of radiological contamination with a direct interface are located close to each other to minimize pipe lengths. Piping carrying radioactive material from these components is routed through shielded pipe chases where practicable. The number of radioactive pipes inside each pipe chases is minimized to reduce the frequency of access into the pipe chase for maintenance activities. This approach is consistent with NRC RG 8.8, Position 2.b(10).

c. Component Access Components have adequate internal space on all sides for easy access for maintenance and inspection. The spacing includes a laydown area or equipment pull area, as well as transport paths for removal or replacement of equipment.

Rigging and lifting equipment is also provided to facilitate the removal, transport, or replacement of equipment or portable shielding during maintenance activities.

This design is consistent with NRC RG 8.8, Position 2.a.

d. Hot tool cribs and hot machine shop Hot tool cribs are located in low-radiation areas adjacent to personnel air locks to minimize the amount of wait time in high-radiation areas, to help prevent the spread of contamination to other plant areas, and to decrease the amount of decontamination work. This design reduces personnel exposure.

The provision of a hot machine shop in the compound building enables personnel to remove equipment and perform maintenance in a lower-radiation area. Access to the hot machine shop is also provided from the truck bay for ease of equipment movement.

e. Staging areas Large staging areas outside the equipment hatch and personnel airlocks allow for prestaging prior to the start of an outage, and provide space for efficient radiation controls for moving equipment in and out of reactor containment building.

12.3-2 Rev. 2

APR1400 DCD TIER 2

f. Personnel decontamination and change areas Personnel decontamination and change areas are adjacent to radiation control area access points. Storage space for protective clothing, respirators, shower and toilet facilities, lockers, and containers for contaminated clothing are provided in these areas.

g. Radiation control area The APR1400 design provides a single point of access into the radiation control areas; however, emergency egress is provided on all elevations. The access area to the radiation control area provides a flexible and adaptable layout to accommodate outage work crews and enhance the availability of immediate interaction with radiation protection personnel stationed at this point.

h. Accessways and entrances to high-radiation areas Labyrinths are provided at the entrances to high-radiation areas to minimize exposure due to scattering and streaming of radiation through entrances and piping penetrations. The scattered dose rate through the passageway and the transmitted dose rate through the shield wall from all contributing sources shall be below the upper limit of the radiation zone specified for each area. This approach is consistent with NRC RG 8.8, Position 2.b(4).

The plant layout design can accommodate removable shields as necessary to provide shielding during normal operation for adjacent corridors. These shields can provide additional shielding during the removal of radioactive components, such as heat exchangers, for maintenance activities. The removable shield is designed to limit the radiation level so as not to exceed the radiation zone boundary in the surrounding areas. Therefore, its density is required to be greater than the permanent concrete shield wall and the gap with the concrete structure is designed to be less than 1/2 with an offset. This approach is consistent with NRC RG 8.8 Position 2.b (2).

High-radiation areas are provided with locked doors to prevent inadvertent access by plant personnel.

12.3-3 Rev. 2

APR1400 DCD TIER 2 High-radiation areas and very-high-radiation areas are properly marked with signage to alert personnel and to further prevent inadvertent access as part of the radiation protection program (see Section 12.5).

Ingress and egress with offset inner walls with curbs are provided for the pumps and valve galleries for maximum radiation attenuation and shielding.

12.3-4 Rev. 2

APR1400 DCD TIER 2 12.3.1.2 Equipment and System Design Features The APR1400 design specifies the use of reliable and simplistic equipment with a minimal number of parts and straightforward operation to reduce the frequency of maintenance and radiation exposure to plant personnel. For the purpose of reducing the radiation field and the exposure to personnel during maintenance, surfaces will be polished in those areas of the plant where this treatment will significantly reduce the dose. The COL applicant is to determine the areas that will require either electro or mechanical polishing (COL 12.3(1)).

The potential areas that may require polishing are as follows:

a. Steam generator channel head, including divider plate

b. Pressurizer shell (pressure boundary portion) except heads and nozzles

c. Reactor vessel closure head and bottom head excluding reactor vessel shells and flange

d. Reactor permanent pool seal (reactor cavity area)

e. Reactor coolant system main piping (pressure boundary portion) excluding surge line and nozzles attached on the piping

f. J-groove weld surface of the reactor vessel closure head

g. Regenerative heat exchanger

h. Spray valves

i. Refueling pool liner

j. Refueling cavity liner

k. Spent fuel pool liner

l. Cask loading pit liner The design characteristics of the equipment used in radioactive systems are as follows:

a. Pumps 12.3-5 Rev. 2

APR1400 DCD TIER 2

1) Pumps and associated piping are flanged to facilitate pump removal to a lower-radiation area for maintenance or repair. Pump internals are also removable. This design approach is consistent with NRC RG 8.8, Position 2.b (9).

2) All pump casings are provided with drain connections to facilitate decontamination. The drain connections are free of internal crevices to minimize accumulation of radioactive corrosion products (crud).

3) Pump seals are easily serviceable without the removal of the entire pump or motor. The seals of reactor coolant pump, centrifugal charging pump, shutdown cooling pump, and safety injection pump are of the cartridge type to facilitate their removal for maintenance or repair.

4) The reactor coolant pump casing is fabricated with austenitic stainless steel cladding and that the finished surface of the cladding is machined to have smooth surfaces to limit the buildup of radioactive contamination on its surface.

b. Ion exchangers (demineralizers)

1) Ion exchangers are designed for complete drainage.

2) Spent resin removal is designed to be done remotely by hydraulic flushing from the vessel to the SWMS.

3) Piping, strainers, and resin screens are flushable so that all spent resin is removed from the ion exchange column and connected piping. This minimizes radiation exposure resulting from spent resin transfer operations.

4) Fresh resin is added remotely from a low-radiation area above the shielded compartment housing the ion exchanger.

5) Ion exchangers are designed with minimal internal crevices to prevent an accumulation of radioactive cruds. This design approach is consistent with NRC RG 8.8, Position 2.f (6).

c. Liquid filters 12.3-6 Rev. 2

APR1400 DCD TIER 2

1) Filter housings are provided with vent connections and are designed for complete drainage.

2) Filter housings are designed with minimal internal crevices to prevent an accumulation of radioactive cruds.

3) Adequate space is provided near cartridge filters to allow for cartridge removal and loading for transport to the solid waste storage area.

4) Filter housings and cartridges are designed to permit the remote removal of spent radioactive filter elements. Cartridge filter seals are an integral part of the filter cartridge and can be removed concurrently with the filter elements.

This design approach is consistent with NRC RG 8.8, Position 2.i (4).

5) Cartridge filter housing closure heads are designed to swing free for the unobstructed removal of the cartridge.

6) Design features to handle spent filters are described in Subsection 11.4.2.

d. Tanks

1) Tanks are designed for complete drainage and are therefore free of internal crevices and pockets. The drain line is connected to the bottom of the tank.

2) Tanks are provided with at least one of the following means of decontaminating the tank internals based on tank contents and radioactivity levels:

a) Ample space to permit decontamination of the tank manway b) Internal spray nozzles on potentially highly contaminated tanks for internal decontamination c) Backflush capability for tank inlet screens

3) Tanks are designed with ellipsoidal or sloped bottoms to facilitate drainage and minimize the accumulation of crud.

4) Tanks are provided with vents to facilitate the removal of potentially radioactive gases. The tanks are vented to the cubicle atmosphere within 12.3-7 Rev. 2

APR1400 DCD TIER 2 close proximity to the respective radioactive building HVAC system intake for collection and treatment of radioactive gases. Tanks are provided with internal filters on the vent lines to prevent the release of liquids

5) Non-pressurized tanks are provided with overflow lines routed to a floor drain pump or other suitable collection point to avoid spillage of radioactive fluids onto the floor or ground. The floor drain system (FDS) is connected to the LWMS for additional processing prior to release to the environment.

6) All tanks containing suspended solids are provided with mixing capabilities to prevent settling of radioactive solids.

e. Valves

1) The following descriptions summarize the valve design features that minimize valve leakage and extend valve design life.

a) A packless valve is used as an isolation valve that is non-modulating, is less than 5.08 cm (2 in.) in diameter, and operates on a weekly interval in the radioactive material processing systems to minimize external leakage from the valves.

b) Modulating valves or valves greater than 5.08 cm (2 in.) in diameter use live loading of the packing by conical spring washers or equivalent means to maintain a compressive force on the packing, in addition to double stem packing with a leakoff between the packings and a graphite lantern ring, where possible.

c) For the valves 5.08 cm (2 in) and under, excluding modulating valves, just double stem packing with a leakoff between the packing and a graphite lantern ring is used. Stem leakage is piped to an appropriate drain sump or tank. All lantern rings used are graphite lantern rings.

All valves with stem leakage are shown on Figures 5.1.2-3, 6.3.2-1, and 9.3.4-1.

d) Valves using stem packing are provided with a backseat capability.

e) Radiation-resistant seals, gaskets, and elastomers are used, when practicable, to extend the design life and reduce maintenance frequency.

12.3-8 Rev. 2

APR1400 DCD TIER 2 f) Valves located in high-radiation areas are equipped with reach rods or motor operators to allow operation from lower radiation zones to minimize radiation exposure.

2) Fully ported valves are used so that valves are fully open and radioactive fluid flows freely. This minimizes internal accumulation of cruds.

3) Valves requiring removal during maintenance and inspection activities are flanged.

4) Internal valve surfaces are designed to be as smooth as possible and free of crevices to minimize the accumulation of crud.

5) Valve wetted parts are made of austenitic stainless steel or a similar corrosion-resistant material.

6) Valves are designed so that they may be repacked without removing the yoke or topworks.

7) Valves for highly radioactive components are located in shielded valve galleries to the extent possible to minimize operator exposure.

8) Check valves are used only where necessary. The type and size of valve selected are compatible with system requirements to reduce disc flutter related wear. Check valves are located and oriented properly in the piping system.

9) The application of metallic bellows and diaphragms is limited to the valves with low-stroke length applications, or infrequent movement in the APR1400 design.

f. Piping and penetrations

1) The field run piping is minimized to the extent practicable to reduce the potential for unintended leakage to the environment.

12.3-9 Rev. 2

APR1400 DCD TIER 2

2) Resin and concentrate piping is designed as follows:

a) The length of pipe runs is minimized by locating the related components as close to each other as possible.

b) Piping is routed through shielded pipe chases whenever possible to minimize the radiation exposure rate in personnel access corridors.

c) Large-diameter piping (greater than 12.7 cm [5 in.] in diameter) is used to minimize the potential for clogging during slurry or resin transfer without violating minimum flow requirements.

d) The number of pipe fittings, such as elbows and tees, is minimized to reduce the potential for radioactive crud accumulation. Where elbows are needed, large-radius bends are used.

e) Low points, deadlegs, elbows, and vertical pipe runs that may cause accumulation of radioactive material are minimized.

f) Pipe runs are sloped and gravitational flow is used where practicable.

g) Crevices on piping internal surfaces are minimized by the use of butt welds instead of socket welds. Socket welds are known to produce crud traps in radioactive systems. The use of butt welds generally results in smoother internal surfaces, reducing crud buildup.

h) Flushing capability is provided to facilitate the decontamination of piping containing radioactive fluids.

i) Penetrations are located so that the source and penetration are not in a direct line and are located high above the floor where practicable. This minimizes the potential for personnel exposure due to radiation streaming pathways.

g. Heat exchangers

1) Heat exchangers are designed for complete drainage and with vents.

12.3-10 Rev. 2

APR1400 DCD TIER 2

2) Internal wetted surfaces are designed to be free of crevices to minimize the potential for the accumulation of radioactive crud on internal surfaces.

3) Corrosion-resistant materials, such as stainless steel, are used to minimize the need for part replacement and reduce the frequency of required maintenance.

h. Reactor vessel head vent A vent nozzle and line are provided on the reactor vessel (RV) head. Use of this design feature allows for a reduction in worker radiation exposure during the head removal process by minimizing the gases discharged directly to the reactor containment building atmosphere while the head is being removed.

i. Reactor coolant system leakage control Exposure from airborne radionuclides to personnel entering the reactor containment building is minimized by controlling the amount of reactor coolant leakage released to the reactor containment building atmosphere. Examples of such controlled leakage are as follows:

1) Pilot-operated safety relief valve (POSRV) leakage is directed to the in-containment refueling water storage tank (IRWST).

2) Valves larger than 5.08 cm (2 in.) in diameter are provided with a double-packed stem with an intermediate graphite lantern ring and a leakoff connection to the reactor drain tank (RDT) with emphasis on the valve stem finish to reduce wear and leakage.

3) Instrumentation is provided to detect abnormal reactor coolant pump (RCP) seal leakage. The RCPs are equipped with two stages of seals, plus a vapor or backup seal, as described in Section 5.4. The vapor or backup seal prevents leakage to the reactor containment building atmosphere and allows sufficient pressure to be maintained to direct the controlled seal leakage to the volume control tank (VCT) and RDT. The vapor seal is designed to withstand full reactor coolant system (RCS) pressure in the event of failure of any or all of the two primary seals.

j. Refueling equipment 12.3-11 Rev. 2

APR1400 DCD TIER 2

1) All spent fuel transfer and storage operations are designed to be conducted underwater to provide reasonable assurance of adequate shielding and limit the contribution of radioactivity in working areas.

2) Equipment is designed to prevent the fuel from being lifted above the minimum safe water depth, thereby limiting personnel exposure and avoiding fuel damage.

3) The equipment design limits the possibility of inadvertent fuel drops, which could cause fuel damage and personnel exposure.

4) The refueling equipment design facilitates the transfer of new and spent fuel at the same time to reduce overall fuel handling time and personnel exposures.

5) Underwater cameras are used to facilitate safe handling through visual control, thus minimizing errors and potential exposures.

6) Portable hydraulic cutters are provided to cut expended control element assemblies (CEAs) and in-core instrumentation (ICI) leads. The cutters allow underwater handling of these items.

7) Equipment is provided to allow for the underwater inspection of fuel elements.

k. Inservice inspection equipment Inspection of the reactor coolant pressure boundary can be done with remote equipment to minimize personnel exposures.

l. Remote instrumentation Systems containing radioactive fluids are designed to be controlled remotely to the extent practicable to minimize personnel exposures.

m. Inservice inspection (ISI) of reactor vessel nozzle welds The design of welds joining the RV nozzle to the reactor coolant pipe permits ISI to be accomplished from inside the RV. Automated equipment, operated remotely, can be used for reactor vessel pressure boundary inspections in this area.

12.3-12 Rev. 2

APR1400 DCD TIER 2 In the event that an ISI of this area is performed from outside of the RV, removable insulation for the RV and reactor coolant piping is used to permit access. These removable sections are lightweight and held in place mainly by quick-application type buckle fasteners. After the necessary panels are removed, remote equipment can be used to perform the required inspections.

n. Blanket-type thermal insulation Blanket-type thermal insulation with hook-and-loop fasteners is selected where needed for components and piping containing radioactive fluids. A metal jacket around the insulation is provided. This jacket is held in place by quick-application type buckle fasteners. This insulation is easily removable to facilitate the performance of inservice weld inspections. This minimizes personnel exposures received during ISI.

o. Electrical service and lighting The APR1400 design provides quality lighting and convenient electrical services to facilitate maintenance and inspection and reduce anticipated personnel exposure.

Reliable lamps of extended service life are used in high-radiation areas whenever possible to minimize the frequency of maintenance/replacement. These features are included in the facility layout design in accordance with the guidance of NRC RG 8.8, Position C.2.i.

p. Spent fuel pool (SFP) and refueling pool decontamination The APR1400 design provides the capability to use high-pressure demineralized water for the decontamination in the SFP and refueling pool. When high pressure demineralized water is used to decontaminate the SFP, and refueling pool, or bulky components, protective covers are required to be used to minimize the spread of contamination. High pressure demineralized water is not to be used in open areas above the pool. Alternative methods of decontamination, such as a strippable coating, may be evaluated by the operator for practicality.

q. Snubbers Mechanical snubbers, rather than hydraulic snubbers, are used in radiation areas to minimize the frequency of required maintenance and inspections.

12.3-13 Rev. 2

APR1400 DCD TIER 2 12.3.1.3 Source Term Control Source term control is an important aspect of the APR1400 design. The following design features reduce the doses received by plant personnel from operations, maintenance, and inspection activities:

a. Fuel performance The APR1400 design features provide reasonable assurance of low primary system source terms, not only because of the extended fuel cycle, but also because of minimized fuel clad leakage based on extremely low fuel clad defects.

b. Corrosion product control The APR1400 design includes design features that reduce corrosion product production in the primary system.

1) Primary system materials The APR1400 design specifies that primary system components are to be fabricated from materials with low corrosion rates and low cobalt impurities (target of 0.02 wt% or less) except where no proven alternative exists. This approach is consistent with NRC RG 8.8, Position 2.e.

Cobalt content for components with a large wetted surface area and an operating temperature of greater than 93 °C (200 °F) is restricted to a maximum of 0.05 wt%. For components in or near the core or components that are expected to release significant quantities of corrosion products, cobalt content (mean value) lower than 0.05 wt% is specified.

The cobalt content for primary system materials are presented in Table 12.3-1.

The presence of antimony in RCP bearings has presented a problem with hot particles in the current generation of nuclear plants. In the APR1400 design, RCP bearings are designed to minimize the amount of antimony.

Nickel-based alloy is used for SG tubes based on industry experience for similar applications in Korean domestic plant.

12.3-14 Rev. 2

APR1400 DCD TIER 2 SG tubes are fabricated to relieve stresses to reduce stress corrosion cracking (SCC), thus reducing the probability that the tubes will plug and reducing radiation exposure to maintenance workers.

Control rod drive materials are specified to be fabricated from low-cobalt alloys to the extent possible.

Cobalt-base hard-facing alloys are used to provide wear resistance only when no proven alternative exists.

2) Primary system chemistry Increasing the primary system water chemistry pH from 6.9 to 7.4 reduces equilibrium corrosion rates and the buildup of activated corrosion products on primary system surfaces. Details are addressed in Subsection 5.2.3.2.1 for the primary system water chemistry pH control. The COL applicant is to establish how the water chemistry pH control reduces radiation fields (COL 12.3(2)).

12.3.1.4 Airborne Contamination Control In the APR1400 design, plant HVAC systems are designed so that airflow is maintained from low-radiation areas to high-radiation areas. This design minimizes the potential for the spread of contamination to low-contamination areas and is consistent with NRC RG 8.8, Position 2.d(1). The reactor containment building is monitored for particulate, iodine, and gaseous activity. Indication of high containment activity automatically initiates containment purge isolation. This approach meets the intent of NRC RG 8.25, Regulatory Position c.2.1 (Reference 5). In addition, the following confinement devices are used to minimize the spread of contamination in accordance with the ALARA principle and NRC RG 4.21 (Reference 4):

a. Leakage inside reactor containment building Leakage inside the reactor containment building is collected in local sumps and transferred to floor drain tanks for processing and release. The sumps are designed to be smaller for more frequent transfer to the LWMS to minimize evaporation and suspension of radioactive particulates inside the reactor containment building. Additional design requirements for the reactor 12.3-15 Rev. 2

APR1400 DCD TIER 2 containment building sump and the ICI sump are provided in Subsection 9.3.3.2.4.d.

b. Ventilation HVAC systems are provided for individual buildings as part of the APR1400 design. Each HVAC system is designed to provide the proper ventilation flow to remove contaminated gases and vapor to levels below the limits of 10 CFR Part 20, Appendix B, Table 1, and to keep building areas, cubicles, walkways, and control areas in the proper environmental conditions (temperatures and air quality). For maximum efficiency, airflow generally is directed from clean/low-contamination areas to higher-contamination areas at a velocity suitable for minimizing entrainment of moisture and particulate. There is also system redundancy to minimize interruption of continuous operation. Exhausted airflows are processed through high-efficiency particulate air (HEPA) filters, charcoal adsorbers, and monitors for radiological contamination before being discharged into the atmosphere. This design approach provides proper airborne contamination control and meets the requirements of NRC RG 8.8, Position 2.d. HVAC systems are addressed in Subsection 9.4.8.

c. Hot machine shop This area provides a dedicated area where maintenance can be performed on radioactive and contaminated equipment. The hot machine shop allows for maintenance and repair activities to be performed in a lower-radiation area with plenty of space for workers to facilitate the efficient completion of the maintenance task.

d. Loop seals Water-filled loop seals are provided in the floor drain system to preclude the flow of contaminated material from one area or floor to another.

12.3.1.5 Equipment Improvements

a. The APR1400 design RCPs incorporate a cartridge type of RCP seal, which is a proven, reliable, and easily replaceable seal design. The replacement of RCP seals is also facilitated by platforms around the RCPs. This design allows the 12.3-16 Rev. 2

APR1400 DCD TIER 2 seal to be removed and repaired outside the crane wall or in another low-radiation area, consistent with NRC RG 8.8, Position 2.b (9). Therefore, the amount of time needed to maintain the RCP seals and the associated occupational exposure is reduced.

b. Steam generator maintenance The APR1400 design includes several features that enhance accessibility to steam generators during maintenance and inspection. These features, described in Subsection 5.4.2, reduce the overall exposure to personnel during maintenance and inspection activities. The features include:

1) Use of automatic or robotic equipment to perform inspection and maintenance activities

2) Adequate pull and laydown areas

3) Platforms to facilitate access to SG parts for maintenance and inspection

4) Handholes to facilitate worker handling of SG parts

5) Adequately sized manways to facilitate easy access for workers performing maintenance and inspection activities

6) Removable insulation to facilitate weld inspection

7) Alloy 690 for tubes to reduce corrosion product production

c. The APR1400 design includes features that are important to achieving ALARA goals. The features are summarized as follows:

1) Equipment selection that is based on reliability, maintainability, and accessibility

2) Component design such as tank design, piping design, and instrument design to minimize particulate deposition

3) System flushing and decontamination capability for systems and components containing radioactive materials 12.3-17 Rev. 2

APR1400 DCD TIER 2

4) Radwaste handling operations are performed remotely, wherever practicable

5) Separation of radiologically contaminated components and the provision of adequate shielding surrounding the rooms/areas containing radioactive components

6) Controlled access to high-radiation areas via locked doors and administrative controls

7) Piping containing radioactive liquid, resins, or gases routed through shielded pipe chases, wherever possible 12.3.1.6 Radiation Zone Designation In order to maintain radiation exposures ALARA throughout the layout and shielding design, the plant is divided into several radiation zones. The zones are used in the development of plant procedures to provide reasonable assurance that maintenance, inspection, and testing activities are performed in the lowest radiation zone possible. The zones indicate maximum dose rates based on the shielding design basis source terms presented in Section 12.2. The radiation zone designations during normal operation are summarized in Table 12.3-2. The radiation zone maps for the auxiliary/reactor containment building are shown in Figures 12.3-1 through 12.3-9. The radiation zone maps for the compound building are shown in Figures 12.3-10 through 12.3-16.

The turbine generator building is generally a non-radiation area. The potentially radioactive area is the condensate polisher room during normal operation and AOOs. The potential radioactivity is from the removal of radiological contaminants by the polishing demineralizers.

Because of the potential for primary-to-secondary (PTS) SG leakage or permeation through SG tubes, the secondary-side plant systems may have a low level of radioactive contamination. The SG blowdown liquid is provided with radiation monitors to detect tube leakage; the condenser vacuum vent effluent radiation monitor also provides an indication of an SG tube leak.

The monitoring systems provide continuous monitoring of the SG performance and while the condensate system and blowdown system process secondary water during SG leakage events; the activities of the spent resins and filter cartridges are continuously monitored and 12.3-18 Rev. 2

APR1400 DCD TIER 2 are not anticipated to be significantly contaminated by PTS leakage. The condensate polisher room and the condenser areas are also provided with area radiation monitors to provide reasonable assurance that personnel radiation exposure is maintained ALARA.

Radiation zone maps for the turbine generator building are presented in Figures 12.3-17 and 12.3-18.

There is a potential for the CCW system to become contaminated through leakage from interfacing radiologically contaminated systems. The CCW system is provided with radiation monitors to ensure that the contamination level is low. As shown in Figure 12.3-19, the CCWS heat exchangers are located in a separate yard building that is designated Radiation Zone 1.

12.3.1.7 General Design Considerations to Keep Post-Accident Exposures ALARA Direct and airborne sources of radiation exposure are considered when determining access provisions to vital areas necessary for the control of the plant. The plant design provides reasonable assurance that personnel exposures meet General Design Criterion (GDC) 19 of 10 CFR Part 50, Appendix A (Reference 6), and NUREG-0737 (Reference 7) guidelines.

To provide reasonable assurance that personnel exposures are kept ALARA, the APR1400 is provided with a post-accident sampling system (described in Subsection 9.3.2) that meets the requirements of NUREG-0737 and NRC RG 1.97 (Reference 8).

12.3.1.8 Post-Accident Radiation Zones Post-accident radiation zone maps are developed in accordance with NRC RG 1.183 (Reference 9) to assess the access throughout the plant following a DBA. The layout of the plant facilities is designed with the intent of keeping occupational doses ALARA during a DBA as well as during normal operation. Post-accident source terms are described in Subsection 12.2.3 and Table 12.2-24.

Continuous access is provided during post-accident conditions, with dose rates less than or equal to 0.15 mSv/hr, to the following vital areas:

a. Main control room

b. Technical support center 12.3-19 Rev. 2

APR1400 DCD TIER 2 Required access to the following vital areas and systems does not exceed the occupational dose limit of 50 mSv:

a. Post-accident sampling system

b. Remote shutdown room and remote control console room

c. Class 1E switchgear room

d. I&C equipment room

e. Access areas outside the CS and SC pump rooms Generic plant emergency procedures are reviewed and the areas listed above are identified as vital areas. The following systems are considered for post-accident access, but do not constitute vital areas:

a. Safety injection system

b. Containment spray system

c. Shutdown cooling system

d. Chemical and volume control system The safety injection system, containment spray system, and shutdown cooling system are all safety-related systems with ESF functions, which are required during design basis accident conditions in order to achieve safe shutdown and mitigate the accident. However, these functions which are required during accident conditions can be monitored and controlled remotely from the MCR and RSR. Therefore, the plant areas housing these system components (with the exception of the area outside of the CS/SC pumps) do not directly require operator access and are therefore not considered vital areas.

The zone limits are summarized for the design basis loss-of-coolant accident (LOCA) in Table 12.3-3. Post-accident radiation zone maps are shown in Figures 12.3-20 through 12.3-51.

12.3.1.9 Vital Area Access 12.3-20 Rev. 2

APR1400 DCD TIER 2 The following descriptions detail the post-accident access routes to the vital areas listed in Subsection 12.3.1.8. Vital area access routes are illustrated by arrows in Figures 12.3-20 through 12.3-51. During normal operating conditions, the radiation control area has a single point of access in the compound building at elevation 30.48 m (100.0 ft). However, during post-accident conditions, emergency access and egress are possible on all elevations if needed.

a. Main control room (MCR)

The MCR is occupied continuously during post-accident situations. The design dose for this area does not exceed 0.15 mSv/hr averaged over 30 days for a TEDE of 50 mSv. The MCR is located in the auxiliary building at El. 47.5 m (156.0 ft).

Access to the MCR is via the compound building. It is generally accessible on all elevations from stairwells and elevators located at 14-15, AH-AJ, and AB-AD.

The locations of the MCR and associated access routes are shown in Figure 12.3-40.

b. Technical support center (TSC)

The TSC is continuously occupied during post-accident situations. The design dose for this area does not exceed 0.15 mSv/hr averaged over 30 days for a TEDE of 50 mSv. The TSC is located in the auxiliary building at El. 47.5 m (156.0 ft).

The TSC is generally accessible on all elevations from stairwells and elevators located at 14-15, AH-AJ, and AB-AD. The locations of the TSC and associated access routes are shown in Figure 12.3-40.

c. Post-accident sampling system (PASS)

The PASS area is irregularly accessed, not continuously occupied, during post-accident situations. The design dose for these areas may not exceed the TEDE of 50 mSv. The post-accident sampling room is located in the auxiliary building at El. 16.8 m (55.0 ft) and the analysis room is in the compound building at El. 25.9 m (85.0 ft). The locations of the post-accident sampling room, analysis room and associated access routes are shown in Figure 12.3-20 and 12.3-12.

d. Remote shutdown room (RSR) and Remote control console room (RCCR) 12.3-21 Rev. 2

APR1400 DCD TIER 2 The RSR and RCCR are accessed to safely shutdown the reactor in the event that the main control room is inaccessible. The design dose for these areas may not exceed the TEDE of 50 mSv. The RSR and RCCR are located in the auxiliary building at El. 41.9 m (137.5 ft). The locations of RSR and RCCR and associated access routes are shown in Figure 12.3-36.

e. Class 1E switchgear room The four (4) class 1E switchgear rooms house the safety-related equipment, which require infrequent access to perform the vital function. The design dose for these areas shall not exceed the total effective dose of 50 mSv. The class 1E switchgear rooms are located in the auxiliary building at an elevation of 23.8 m (78.0 ft). The locations of the class 1E switchgear rooms and associated access routes are shown in Figure 12.3-24.

f. I&C equipment room The four (4) I&C equipment rooms house the safety-related instrument and control systems, which require infrequent access to perform the vital function. The design dose for these areas shall not exceed TEDE of 50 mSv. The I&C equipment rooms are located in the auxiliary building at an elevation of 47.5 m (156.0 ft). The locations of the class 1E switchgear rooms and associated access routes are shown in Figure 12.3-40.

g. Access areas outside the CS and SC pump rooms The areas outside the CS and SC pump rooms are irregularly accessed, when the manual actuation is required after post-accident situations. The manual actuation involves the opening and closing of valves for the interchanging of the CS and SC pumps and the opening or closing of valves for the connections of these two systems to the IRWST. These configurations may be implemented as back-up mitigation measures depending on the changing post-accident conditions. The design dose for these areas shall not exceed TEDE of 50 mSv. These areas are located in the auxiliary building at an elevation of 16.8 m (55.0 ft). The locations of these areas and associated access routes are shown in Figure 12.3-20.

12.3-22 Rev. 2

APR1400 DCD TIER 2 12.3.2 Shielding The shielding design is based on the source terms, design dose rates, and established design criteria in Subsection 12.2.1 and is in accordance with the calculation method and guidance in NRC RG 1.69 (Reference 10) and NRC RG 8.8.

12.3.2.1 General Shielding Design Criteria Shield walls are provided around components that contain and handle radioactive materials for worker safety and to maintain radiation doses ALARA. The wall thicknesses listed in Table 12.3-4 are based on the shielding basis source terms of the component, the design dose rate, and the shielding material. This approach is consistent with NRC RG 8.8.

The shielding design for the MCR and the primary shielding in the reactor containment building is safety-related. The shielding for the MCR meets the requirements of 10 CFR Part 50, Appendix A, GDC 19. Within the reactor containment building, the pressurizer is separated from the operating floor by a shield wall with a minimum thickness of 33 inches.

The steam generators, reactor coolant pumps, and associated RCS piping are also separated from the operating floor and annulus area by the secondary shield wall which has a minimum concrete thickness of 48 inches.

Radiation protection of personnel, equipment, and materials is dependent primarily on the adequacy of the design of the plant shielding. Radiation shielding has the passive protection function of radiation attenuation and consists of materials placed between radiation sources and plant personnel as well as the public. The shielding system is designed and constructed to provide reasonable assurance that the station can be operated and maintained so that the resultant radiation levels and doses are below the limits of 10 CFR Part 20 and are ALARA. Design dose rate limits to achieve this objective are addressed in Subsection 12.3.1 and listed in Table 12.3-2.

Shielding is designed for the plant lifetime and under the various radiation source and environmental conditions associated with normal operation, AOOs, and DBA conditions identified in Chapter 15 and summarized as follows:

a. Plant conditions

1) Normal operating condition 12.3-23 Rev. 2

APR1400 DCD TIER 2 For shielding design, normal station operating conditions are considered to include conditions generally known as AOOs. Two modes of normal station operation are:

a) Normal power operation of the reactor b) AOOs (startup, shutdown, and refueling of the reactor)

Shielding is designed to provide a protective function under such conditions. Normal operation zone designations are provided in Table 12.3-2.

2) Accident conditions Shielding provides protection to plant operating personnel and the general public under postulated DBA conditions as defined in Chapter 15.

a) MCR habitability The MCR and associated areas are designed according to the design requirements in 10 CFR Part 50, Appendix A, GDC 19, and are shielded so that, after a postulated DBA, radiation exposure in the MCR for the duration of the accident does not exceed the TEDE of 50 mSv, including dose contributions from ingress and egress of the MCR.

The radiation shielding protecting the MCR and associated areas is designed based on the anticipated radiation environment resulting from the postulated DBA.

b) Direct offsite doses Adequate shielding is provided to limit the total effective dose equivalent not to exceed the limits specified in 10 CFR 52.47 (Reference31) and Chapter 15 of the Standard Review Plan (SRP) (Reference 12). The shielding is designed adequately that the radiation dose to an individual located at the exclusion area boundary for a duration of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during a postulated DBA does not exceed 25 rem TEDE.

b. Seismic and safety classification 12.3-24 Rev. 2

APR1400 DCD TIER 2 Structural walls are designed to meet seismic category requirements. Structural walls are designed as seismic Category I, II, or III depending on the particular design requirements other than the radiation protection requirements such as structural integrity and load-bearing capacity.

The primary shield, the shield walls for the MCR, and the shield walls for the SFP are examples of shield walls that are designed as seismic Category I.

c. Protection of equipment and structures Adequate shielding is provided for the following purposes:

1) To limit radiation heating of structural concrete

2) To reduce neutron activation of equipment

3) To limit the radiation dose to equipment and materials

d. Maintenance, inspection, and testing considerations Adequate shielding is provided to provide reasonable assurance of safe personnel access and sufficient stay time near areas containing radioactive equipment for maintenance, inspection, and testing activities.

e. Additional requirements The other shielding systems functional requirements generally depend on the location of the shield and the access requirements to or from the equipment or areas within the shield walls. Thus, access to an area may be through the shield itself such as through removable shield walls.

12.3.2.2 Shielding Analysis Calculations to determine the adequacy of the shielding design are based on the source strengths described in Subsection 12.2.1 and the methods described below. Dose points are selected for analysis inside and outside cubicles containing radioactive equipment.

Skyshine from the cubicles is negligible because cubicles containing radioactive material are shielded overhead.

12.3-25 Rev. 2

APR1400 DCD TIER 2 The only major source generating radionuclides is the reactor core at full power. The codes ANISN (Reference 13) and MCNP (Reference 14) are used to verify the effectiveness of the primary shield and evaluate the streaming of neutron and gamma radiation from the reactor vessel. Sources of gamma radiation are distributed throughout the reactor containment building and nuclear island. The codes MICROSHIELD (Reference 15) and RUNT-G (Reference 16) are used to verify gamma source shielding.

The shielding analyses are based on a concrete density of 2.242 g/cm3 (140 lbs/ft3). The following sequence typifies a gamma source shielding analysis:

a. Determine the concentration of each principal nuclide in the source medium.

b. Adjust the concentration to account for issues such as accumulation, dilution, decay, and removal.

c. Convert the resulting concentrations into gamma source strength.

d. Select a model or combination of models to represent the physical geometry of the source container(s) and all shields present.

e. Assemble the necessary data on attenuation properties of the source and the shield materials.

f. Perform the calculation for the desired dose point location and tabulate the results for comparison with design objective dose rates.

Computer codes necessary to perform the above analysis meet the related regulatory requirements and industry standards. Computer codes used in shielding analysis are described below:

a. ANISN ANISN is a program used to perform the shielding and heat generation rate calculations for the reactor primary shield inside of containment. ANISN is a multigroup, one-dimensional, discrete-ordinates transport program that solves the one-dimensional Boltzmann transport equation for neutrons and gamma rays in slab, sphere, or cylindrical geometry.

b. MCNP 12.3-26 Rev. 2

APR1400 DCD TIER 2 MCNP is a continuous-energy, generalized-geometry, time-dependent, coupled neutron-photon-electron Monte Carlo transport code system. MCNP is widely used for verification of shielding design adequacy in dose rate reduction because it provides more realistic modeling capabilities and more reliable calculation results than other programs. MCNP is used in shielding calculations as described in Section C.1.a of NRC RG 1.69. This code is used to evaluate the dose rate impact on the operating floor in the containment building as a result of the streaming of neutron and gamma radiation from the reactor vessel through the reactor cavity.

c. MICROSHIELD MICROSHIELD is a comprehensive photon/gamma ray shielding and dose assessment program that is widely used for designing shields, estimating source strength from radiation measurements, and teaching shielding principles.

MICROSHIELD has 16 geometries that accommodate offset does points and as many as 10 standard shields, plus source self-shielding and cylinder cladding.

This code uses library data for radionuclides, attenuation, and buildup and the dose conversion reflects standard data from International Commission on Radiological Protection (ICRP) Publications 38 and 107, as well as ANSI/ANS standards and Radiation Safety Information Computational Center (RSICC) publications. This code is used to determine the radiation zoning and minimum shield thickness requirements of the cubicles containing radioactive sources.

d. RUNT-G RUNT-G uses a modified version of the multiple-compartment model from the RACER-II program to calculate time-dependent radionuclide activities, the source geometry models from the Gaussian quadrature version of the ISOSHLD (Reference 17) program to calculate the resulting time-dependent dose rates, and a simple trapezoidal integration scheme to determine cumulative doses. All of the containment model parameters used in RACER-II and shielding geometries available in ISOSHLD are included in RUNT-G. This code is used to determine the post-accident radiation shielding requirements to provide adequate accessible conditions and to calculate the total integrated dose (TID) for equipment qualification.

12.3-27 Rev. 2

APR1400 DCD TIER 2 12.3.2.3 Shielding Design The plant shielding is designed to achieve the radiation zones designated in Tables 12.3-2 and 12.3-3 for normal operation and post-accident conditions, respectively.

The shielding design thicknesses, as presented in Table 12.3-4, are based on the use of ordinary concrete as the shield material.

In accordance with Standard Review Plan 12.3-4, Sections II.1 and III.3 and Appendix B, and RG 8.38 (Reference 18) C.1.6(3), adequate shielding or operational precautions are provided for all accessible plant areas capable of radiation levels greater than 1 Gy/hr.

Areas with the potential for radiation greater than 1 Gy/hr are listed in Table 12.3-5.

Transient sources greater than 1 Gy/hr are considered in the shielding design to provide reasonable assurance that adequate shielding is provided. One such source is a spent fuel assembly. During transfer of a spent fuel assembly through the fuel transfer tube, adjacent areas may have elevated radiation levels. Streaming from this source up through the joint between the reactor containment building and the auxiliary building has been a concern for the current generation of nuclear plants. The APR1400 design uses connected building structures to reduce the potential for streaming. In addition, sufficient concrete shielding is provided to maintain radiation levels in adjacent areas ALARA during spent fuel transfer.

This permits personnel to perform maintenance and inspection activities in a lower-radiation area and reduces the potential for high-radiation levels adversely affecting refueling outage schedules. An inspection area is provided for the fuel transfer tube.

Access control is provided by the personnel airlock through the reactor containment building. Limited access to the annulus areas within the reactor containment building is allowed during power operation with a tightly controlled occupancy time. Access to these areas is restricted through administrative controls as well as a physical lock on the door.

Sufficient shielding provides reasonable assurance that the areas adjacent to the spent fuel transfer tube are accessible and expected radiation zones are consistent with those in Figure 12.3-52 during transfer of a spent fuel assembly. The shielding design of the fuel transfer tube is based on the 100 hr decayed spent fuel source strengths provided in Table 12.2-9.

The gamma source strengths given in units of [MeV/W-sec] are converted to [photons/sec]

by multiplying the gamma source strength values by the thermal power per fuel assembly in

[W] and dividing by the source energy in [MeV]. Then, the shielding source term is determined by multiplying this calculated value by the radial power peaking factor of 1.55 12.3-28 Rev. 2

APR1400 DCD TIER 2 and by the number of fuel assemblies transferred through the transfer tube, which is two (2).

It should be noted that the rod radial power factor of 1.55 used for the shielding analysis is the ratio of the average power per unit length, while the peak assembly power of 1.2353 is the maximum assembly power during the first cycle for steady-state.

The spent fuel pool is designed to comply with the following shielding design criteria: (a) to provide sufficient shielding so as to meet the radiation zoning requirements adjacent to the SFP and (b) to maintain the dose rate to the spent fuel handling operator less than 0.025 mSv/hr as specified in ANSI/ANS-57.1-1992.

In determining the shielding requirements around the SFP, the source terms for the spent fuel assemblies seated in the SFP storage racks and for the SFP water are considered. To add conservatism to the design, the SFP is assumed to be filled with 1,696 spent fuel assemblies decayed for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, for which the source term is provided in Table 12.2-9.

The radial power peaking factor is not considered. The source region is assumed to be homogeneously mixed with UO2, Zircalory-4 and H2O in the shielding calculations, while the shielding effects for the SFP storage racks and the SFP liner are not taken into account.

For the source terms in the SFP water, the design basis specific activities provided in Table 12.2-17 are used. By using this conservative approach in the shielding calculations, it is ensured that the minimum shielding requirements around the SFP provided in Table 12.3-4 meet the radiation zoning designations presented in Figures 12.3-3 through 12.3-5.

The dose rate to an operator on the refueling platform is determined by the dose rate value corresponding to the axial distance (or water depth) from top of active fuel to the refueling pool water level after 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> decay in Figure 12.2-1. Since mechanical stop in both the refueling machine and spent fuel handling machine restrict withdrawal of the spent fuel assemblies above the minimum safe water cover depth of 274.32 cm (9 ft) from top of active fuel, an operator on the refueling platform will not be exposed to the radiation dose limits in the work area when the shielding of the fuel handling equipment is taken into account.

For shielding of the cask loading pit, it is assumed that a single spent fuel assembly is being loaded into the cask within the cask loading pit. The source term of the spent fuel assembly is conservatively assumed to be decayed for 1,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> after reactor shutdown.

This source term is also provided in Table 12.2-9 and the gamma source strengths used for the cask loading pit shielding are derived using the same methods as those for the fuel transfer tube shielding except that the number of fuel assemblies is one.

12.3-29 Rev. 2

APR1400 DCD TIER 2 Typically, pipe chases do not need to be accessed frequently. The APR1400 design minimizes locating components such as valves in pipe chases to minimize plant personnel access to pipe chases and to reduce the potential for radiation exposure. When access is needed, radiation protection personnel conduct a survey of the area to determine the strength and location of radiation sources within the pipe chase. Temporary shielding is used to minimize personnel exposure. If the primary source of radiation in the pipe chase is spent resin or slurry transfer piping, precautions are taken by operating personnel to provide reasonable assurance that no spent resin is transferred while personnel are in the pipe chase. The resin transfer lines are also provided with a flushing capability to minimize the potential for hot spots in the piping.

Shielding thicknesses and radiation zones in the pipe chases and valve rooms are determined based on the number and characteristics of the pipes in the corresponding areas.

The dose rates for different kinds of pipes are determined using the corresponding source terms in each pipe. The length of each pipe is assumed to be 20 feet because the dose rate at 1 foot away from the pipe is maximized when the pipe is 20 feet long. If there are multiple number of pipes in a pipe chase or a valve room, the resultant dose rate in the cubicle and the required minimum shield thicknesses are determined by summing the dose contributions of each pipe.

Two layers of shield plugs are installed around the reactor vessel. The shield plugs are made of ordinary concrete and are designed to reduce the neutron and gamma streaming from the reactor as low as possible such that the dose rate on the operating floor is minimized. Adequate air cooling is provided to the area between the reactor and the shield plugs and the primary biological shield to ensure that the heat generated from the reactor does not cause any functional deterioration of the concrete with respect to the shielding and structural integrity.

The APR1400 ICI design adopts the bottom mounted In-Core Instrument system that has only fixed type detectors for normal operation. During refueling operation, all ICIs are withdrawn from the core up to the full length of 24 feet when the reactor area is flooded with refueling water. The highly irradiated portion of the ICIs still remains inside the ICI guide tubes which are filled with refueling water at the lower head area of the reactor vessel.

The withdrawal and disposal operation of ICI is performed remotely and the operators are shielded by the pool of water as well as the distance between the reactor vessel and the CEACP; and administrative procedures are used to monitor radiation level to insure personnel safety and radiation is ALARA.

12.3-30 Rev. 2

APR1400 DCD TIER 2 Components that handle a significant amount of radioactive materials, such as LWMS floor drain tanks and equipment waste tanks, are located in shielded cubicles separated from the pump and valve galleries that are provided with labyrinths for access to the galleries. This design approach minimizes radiation streaming and scattering but permits inspection and maintenance access and removal of smaller items such as pumps, valves, and instruments for repair in lower-radiation areas. This design approach meets the requirements of NRC RG 8.8 2.b(4). The doors or hatches being relied on to maintain doses within the radiation zone designations provided in the Chapter 12 radiation zone figures are as follows:

a. Personnel Air Lock between Containment Annulus Area (100-C01) and Personnel Air Lock Entrance (100-A14A)

b. Personnel Air Lock between Operating Area (156-C01) and Containment Entrance Area (156-A04B)

c. Equipment Hatch between Operating Area (156-C01) and Equipment Hatch Access Room (156-A10A)

d. Door between Equipment Hatch Access Room (156-A10A) and the building exterior

e. Doors between Truck Bay (100-P08) and the building exterior The COL applicant is to provide the material composition and shielding properties of these doors/hatches, and these thicknesses equivalent to the minimum required concrete shield thicknesses. Also, the COL applicant is to provide the service life of these doors/hatches and perform periodic inservice inspection and maintenance for these doors/hatches to provide reasonable assurance of functionality throughout the life of the plant (COL 12.3(3)).

The plant shielding is designed not only to maintain personnel occupational exposure ALARA, but also to maintain exposure to the general public ALARA.

The APR1400 shielding design has target dose rates that are below the limits for radiation zone designations provided in Table 12.3-2 to provide a sufficient margin in maintaining radiation exposure to plant personnel and the public ALARA. 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(4))

12.3.2.4 Access Control to High Radiation and Very High Radiation Areas 12.3-31 Rev. 2

APR1400 DCD TIER 2 The high radiation and very high radiation areas, areas potentially greater than 1 Gy/hr and 5Gy/hr, respectively, as identified in Table 12.3-5, which are located in the containment building have multiple features of access control to prevent inadvertent radiation exposure to plant personnel. These very high radiation areas include the ICI cavity, the hold-up volume tank, the core debris chamber, the reactor cavity, the steam generator cavity, and the reactor drain tank room. Access to the containment building is strictly controlled and built-in design features to prevent inadvertent access include a secure air lock as the only point of entry for personnel, the door to which is locked and equipped with a security alarm.

In addition to the access control provided at the point of entry into the containment building, separate barriers with individual locked doors are provided for each of these very high radiation areas in accordance with the guidance of RG 8.38 (Reference 18).

The high radiation areas on Elevations 78 and 86 of the auxiliary building are located within a block where thick concrete walls are provided as shielding to the surrounding areas.

There are no doors provided to allow access to the high radiation cubicles within this block.

These cubicles include the pre-holdup ion exchanger pit, the purification ion exchanger pit, the purification filter pit, and the filter area. This block of filters and ion exchangers can only be accessed from the elevation 100 level via manway, which are locked at all times and are further under administrative controls to prevent unauthorized access. Also on Elevations 100 and 120 of the auxiliary building is the volume control tank cubicle, which is a potentially high radiation area. This cubicle, which is not normally accessed by personnel, is locked and can only be opened by key from the outside.

The areas listed in Table 12.3-5 at the Elevation 120 level of the auxiliary building, which are high radiation areas during refueling operations, include the transfer tube inspection area, the cask loading pit, the refueling canal, and the spent fuel pool. The cask loading pit and refueling canal, and the spent fuel pool do not allow for inadvertent personnel access as these areas do not have an entrance for personal entry, and since the transfer tube access area is locked normally, the transfer tube inspection area cannot be accessed.

The areas listed in Table 12.3-5 as high and very high radiation areas within the compound building are all provided with access control in the form of locked doors. These rooms are provided with a latch bolt operated by key from the outside or by a rotating inside knob/lever. The two exceptions to this form of access control are the hot pipe way on Elevation 77 and the charcoal delay bed room. The hot pipe way and the charcoal delay bed room are not provided with a door for personnel access. The only accesses to these areas are via the hatches provided on Elevation 85 and 120, respectively. Since these 12.3-32 Rev. 2

APR1400 DCD TIER 2 hatches are intended for maintenances or equipment removal, and are equipped with heavy concrete blocks, unauthorized access is not possible.

12.3.3 Ventilation The APR1400 design provides shielding for ventilation components including cubicle ventilation for cubicles that contain components with radiological contamination, filters in the HVAC air cleaning units (ACUs) that remove radiological contaminants from the air streams before release, and ducts carrying radiologically contaminated vent streams to the maximum extent practicable. The shielding design for the ventilation exhaust treatment components are determined using adjusted design basis airborne radiological source terms presented in Table 11.3-1 by multiplication factors derived from the ratio of the activity increase of each nuclide between design basis and expected source terms. The source for each of the ACU filters is calculated based on the buildup of the removed radionuclides during its period of operation up to one year. The dose rates determined the individual filters are summed to determine the overall dose rate for the ACU(s) for determination of radiation zoning and the minimum shield wall thicknesses required. HVAC duct chases are provided with structural walls of 18-inch concrete walls in the auxiliary building and compound building, for shielding purpose.

The spread of airborne contamination within the plant is minimized by the design of the plant HVAC systems to provide airflow from areas of lower potential for airborne contamination to areas of greater potential for airborne contamination. For building compartments with the potential for contamination, the exhaust from the areas is designed with pressure and flow balances to minimize the amount of uncontrolled exfiltration from these areas. These design features provide reasonable assurance that the average concentration of radioactive material in the air in the areas that are normally occupied is less than the small fraction of DAC prescribed in 10 CFR Part 20 Appendix B. Therefore, personnel exposure due to inhalation of and contact with airborne contamination is maintained ALARA.

Airborne radiation monitoring is provided for areas that are normally occupied and have a significant potential for airborne contamination. The monitors can detect the time-integrated change of the airborne radioactivity within 10 DAC-hours for the most limiting particulate and iodine species in each area.

12.3-33 Rev. 2

APR1400 DCD TIER 2 Airborne radiation monitors are described further in Section 11.5. The locations of the process effluent radiation monitors are shown in Figure 11.5-2. The airborne radiation monitors are located upstream of the filters within the HVAC ventilation systems.

HVAC systems are described in Section 9.4.

12.3.4 Area Radiation and Airborne Radioactivity Monitoring Instrumentation The area radiation monitoring system (ARMS) supplements the personnel and area radiation survey provisions of the plant health physics program described in Section 12.5 and provides reasonable assurance of conformance with the personnel radiation protection requirements of 10 CFR 20, 10 CFR Part 50, 10 CFR Part 70 (Reference 19); the guidelines of NRC RGs 1.21 (Reference 20), 1.97, 8.2 (Reference 21), 8.25 (Reference 5), and 8.8 (Reference 1); and American National Standards Institute (ANSI) N13.1-1999 (Reference

22) and Institute of Electrical and Electronics Engineers (IEEE) Std. 497-2002 (Reference 23). The ARMS is in conformance with ANSI/ANS HPSSC-6.8.1 (Reference 24).

The process and effluent radiation monitoring system and sampling systems are described in Section 11.5.

Portable instruments are used and the associated training and procedures are provided to accurately determine the airborne iodine concentration in areas within the facility where plant personnel could be present during an accident in accordance with the requirements of 10 CFR 50.34(f)(2)(xxvii) (Reference 11) and the criteria in Item III.D.3.3 of NUREG-0737. Portable instruments are also used as needed during normal operation in accordance with the guidelines of RG 8.8. 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(5)).

With regard to the criticality accident monitoring, the requirements in 10 CFR 50.68(b)

(Reference 25) are followed to prevent criticality as described in Subsection 9.1.1.

12.3.4.1 Area Radiation Monitoring System 12.3-34 Rev. 2

APR1400 DCD TIER 2 12.3.4.1.1 Design Objective The ARMS monitors the radiation levels in selected areas throughout the plant. Most area monitors are designed to warn operators and station personnel through visible and audible alarms when unusual radiological events occur. Some area monitors are designed to monitor the post-accident radiation level in areas where access to equipment that is important to safety may be necessary. These post-accident radiation monitors, shown in Section 7.5 and Table 7.5-1, are designed in accordance with NRC RG 1.97, Rev. 4 and NRC RG 1.206 (Reference 26).

Area radiation monitors have local visible and audible alarms. An additional visual indication lamp may be provided in high-noise areas if needed to provide reasonable assurance of prompt recognition by nearby personnel during high-radiation conditions.

12.3.4.1.2 Location of Area Radiation Monitors Area radiation monitoring equipment is used to alert operators and station personnel of abnormally high-radiation conditions in an area and protect personnel from possible overexposure. The locations of the area monitors are based on the potential for significant radiation levels in an area. Area monitors are also located in areas where accident access to safety-related equipment may be required during post-accident conditions. Area radiation monitors are also used in special process applications. For example, area radiation monitors located next to the main steam lines are used to monitor for an SG tube leak or rupture and high-range area detectors are used to estimate the accident containment airborne activity and primary coolant activity.

12.3.4.1.3 General System Description Area radiation monitors consist of electronics/displays (RT) and Geiger-Mueller (G-M) tubes or ionization chambers for detecting gamma radiation (RE). Each monitor using RT may be configured with one or more detectors and may cover multiple areas. Selection of detectors is based on the range needed for the particular monitoring application. Some areas may require extended or high-range detector configurations to cover special operational or post-accident monitoring functions.

Radiation level signals and alarms and operation status alarms are generated by each RT for local alarm capability and transmittal to the information processing system (IPS), qualified indication and alarm system (QIAS), and other interfacing systems. The signals and 12.3-35 Rev. 2

APR1400 DCD TIER 2 alarms are recorded and can be retrieved by the operators using IPS. Radiation level signals are transmitted from the RT via digital communication ports and analog outputs.

Alarm relay contacts are provided to actuate alarms for radiation, high radiation, and operation status.

Table 12.3-6 lists the area radiation monitors and specifies the range, electrical class, and seismic category of each monitor.

Area monitors are located based on the expected frequency of access, occupancy time, and expected and potential radiation levels in plant work areas.

a. Areas that are typically high-radiation areas but require little or no access (e.g.,

pipe chases) are not provided with an area radiation monitor.

b. Areas that typically have a high frequency of access and are normally low-radiation areas, but are potentially high-radiation areas, are provided with an area radiation monitor. In accident conditions, areas such as corridors outside personnel and equipment hatches may become very-high-radiation areas and are therefore provided with an area radiation monitor to provide reasonable assurance of worker safety.

During accident conditions, plant personnel are evacuated from the radiation control area. A radiation protection technician then conducts analyses using portable radiation monitors to determine the optimal routes to vital areas through the radiation control area so that personnel exposure is minimized. The radiation protection technician typically escorts maintenance personnel and operators into the radiation control area and continues to monitor radiation levels using portable radiation monitors. The area monitors located in the plant provide an audible and visual alarm if high radiation levels are detected.

12.3.4.1.4 Redundancy, Diversity, and Independence The ARMS, which performs safety functions, is designed to meet the single-failure criteria, separation criteria, segregation criteria, and environmental and seismic qualification requirements in accordance with the requirements of IEEE Std. 323 (Reference 28), IEEE Std. 344 (Reference 29), and IEEE Std. 603 (Reference 30). The balance-of-plant engineered safety features actuation system (BOP ESFAS) consists of the ARMS and the 12.3-36 Rev. 2

APR1400 DCD TIER 2 process and effluent radiation monitoring system described in Section 11.5. The safety evaluation for the ESF system of these monitors is described in Section 7.3.

12.3.4.1.5 ARMS The ARMS monitors presented in Table 12.3-6 are categorized into subsystems used for process monitoring functions or other monitoring applications. The location and flow diagrams of the monitors are shown in Figure 11.5-1.

The ARMS monitors are as follows:

a. Safety-related area monitors (RE-231A, 232B, 241A, 242B, 233A, and 234B)

Containment operating area monitors (RE-231A, 231B) and spent fuel pool area monitors (RE-241A, 241B) are installed as safety-related area monitors for actuating engineered safety features. These monitors perform additional safety functions that generate containment purge isolation and fuel area emergency ventilation actuating signals. The fuel area normal ventilation system is isolated, and the emergency ventilation system is initiated by the fuel area emergency ventilation actuating signal. Containment purge isolation actuating signal (CPIAS) isolates the containment purge system. These monitors are accident monitoring instrumentation (AMI), type C, and also listed on Table 7.5-1.

The containment upper operating area monitors (RE-233A, 234B) consist of physically independent and electrically separated detectors located inside the containment away from the influence of the reactor coolant system to measure high-range gamma radiation. This monitor gives operators a seismically and environmentally qualified indication of containment airborne activity. These monitors conform with the requirements of 10 CFR 50.34(f)(2)(xvii) and the criteria in Attachment 3 to Item II.F.1 of NUREG-0737 and NRC RG 1.97.

The monitors (RE-233A, RE-234B) transmit the radiation signals to the licensing entity via emergency response data system (ERDS) link.

One of the spent fuel pool area monitors is located on a wall and the other near the SFP bridge area.

The containment operating area monitors (RE-231A and 232B) are located near El.

160 directly above the refueling pool. The containment upper operating area 12.3-37 Rev. 2

APR1400 DCD TIER 2 monitors (RE-233A and 234B) are azimuthally 180° apart and located just below the containment polar crane rail support girder (near El. 230). Thus, the two monitors have a wide open, unobstructed view of the entire containment free air volume. The location of RE-231A, RE-232B, RE- 233A, and RE-234B is shown in Figure 11.5-2A.

The location of RT-231A and RT-233A is shown in Figure 11.5-2BB.

The location of RT-232B and RT-234B is shown in Figure 11.5-2M.

The location of RE/RT-241A and RE/RT-242B is shown in Figure 11.5-2O.

b. Non-safety-related area monitors (RE-205, 235, 236, 237, 238, 245, 257, 275, 279, 284, 285, 288, 289, 292, and 293)

A post-accident primary sampling room area monitor (PR-RE-205) is provided in the auxiliary building and a normal primary sampling room area monitor (PR-RE-285) is provided in the compound building.

Two area monitors (RE-237, 238) are provided in the main steam/feedwater piping penetration area and one area monitor (RE-236) is provided in the containment personnel access hatch.

An area monitor is provided for the main control room (RE-275) and Technical Support Center (PR-RE-279) to measure gross gamma dose rates.

Other areas where an area monitor is provided are the in-core instrument area (RE-235), hot machine shop area (RE-293), new fuel storage area (RE-245),

radiochemistry lab area (RE-257), waste drum storage area (RE-292), compound building truck bay area (RE-288, 289), and dry active waste storage area (RE-284) in compound building.

The location of RE/RT-205 is shown in Figure 11.5-2D.

The in-core instrument (ICI) area monitor, RE-235, is a radiation monitor for the ICI seal table area. The seal table is located at Elevation 130-0 near the integrated head assembly lift rig. The monitor is located above the water level (concrete floor level) of 156-0 and the monitor is at 6-6 off the 156 level because the seal table area is flooded and is submerged under water during 12.3-38 Rev. 2

APR1400 DCD TIER 2 refueling operations. RE-235 is located directly overlooking the seal table.

Containment personnel access hatch area monitor, RE-236 is located on the containment wall near the personnel access hatch at Elevation 160-6. The location of RE-235 and RE-236 is shown in Figure 11.5-2A.

The location of RT-235, RT-236, and RE/RT-245 is shown in Figure 11.5-2O.

The location of RE/RT-237 is shown in Figure 11.5-2L.

The location of RE/RT-238 is shown in Figure 11.5-2K.

The location of RE/RT-257 and RE/RT-285 is shown in Figure 11.5-2S.

The location of RE/RT-275 and RE/RT-279 is shown in Figure 11.5-2N.

The location of RE/RT-284, RE-288, RT-288, RE-289, RT-289, RE-292, and RT-292 is shown in Figure 11.5-2T.

The location of RE-293 and RT-293 is shown in Figure 11.5-2U.

c. Alarm location and function of the area monitors The RT of the area radiation monitor is equipped with alarm components such as beacon light (RL) as a visual alert and a horn alarm (RAH) as an audible alarm.

For the containment operating area and the containment upper operating area monitors (RE-231A, 232B, 233A, and 234B), the local alarms are located at the locations of the RE and at the location of the RT.

Common combined RL/RAHs for RE-231A, 232B, 233A, and 234B are located in the containment operating area (RL/RAH-1) and another one (RL/RAH-2) outside the containment building near the personnel access hatch as shown in Figures 11.5-2A and 11.5-2O. RL/RAH-1 and RL/RAH-2 are in addition to an individual RL/RAH for each RT (RT-231A, 232B, 233A, and 234B) in the Auxiliary Building as shown in Figures 11.5-2M and 11.5-2BB. The common inboard alarm alerts the operation and maintenance crew in the containment operating area to exit the containment and also warns the crew outside to limit the access to the containment building.

12.3-39 Rev. 2

APR1400 DCD TIER 2 For the in-core instrument area monitor (RE-235), two local alarms are provided, one located at the location of the RE and the other one at the location of the RT outside the personnel access hatch as shown in Figures 11.5-2A and 11.5-2O.

The local alarm in the in-core instrument seal table area alerts the crew to exit the area and also warns the crew outside to limit access to the area inside containment.

For the containment personnel access hatch area monitor (RE-236), the local alarm is located at the location of the RT outside the personnel access hatch as shown in Figures 11.5-2A and 11.5-2O. The local alarm of the RT alerts the crew outside about high radiation condition inside the personnel access hatch. This alarm at the outboard side of the access hatch alerts the plant crew prior to entering the containment through the access hatch. Since the containment evacuation alarms provided by RL/RAH-1 from RE-231A, 232B, 233A and 234B, a local alarm near RE-235 is not necessary.

For the hot machine shop area monitor (RE-293), the local alarm is located at the location of the RE and at the location of the RT as shown in Figure 11.5-2U.

For the truck bay area monitor (RE-288 and RE-289) and the waste drum storage area monitor (RE-292), the local alarm is located at the location of the RE and at the location of the RT as shown in Figure 11.5-2T.

For the Post-accident primary sample room area monitor (RE-205), the Main steam and FW containment piping penetration area monitors (RE-237 and RE-238), the Spent fuel pool area monitor (RE-241A and RE-242B), the New fuel storage area monitor (RE-245), the Radiochemistry lab area monitor (RE-257), the Main control room area monitor (RE- 275), the TSC area monitor (RE-279), the Compound building dry active waste storage area monitor (RE-284), and the Normal primary sample room area monitor (RE-285), each associated local alarm is located at each location of the RT which is installed adjacent to the RE. The local alarm alerts the crew inside about high radiation condition in the area.

Some of these monitors are located in areas where the areas are continuously manned and occupied where radioactive materials are handled, stored or processed.

Some are located in an open area where distinguishing the inside or outside of an area is immaterial. All these areas, therefore, would not need to distinguish the outside/inside of an area for the purpose of warning the crews before entry into the area.

12.3-40 Rev. 2

APR1400 DCD TIER 2 12.3.4.1.6 Range and Alarm Setpoints The ranges of the ARMS are shown on Table 12.3-6. Alarm setpoints for safety-related monitors are determined by plant procedures and the Offsite Dose Calculation Manual (ODCM). The setpoint methodology includes the relationship between the analytical limit, setpoint, and channel uncertainty. The setpoint methodology also provides channel uncertainty calculations associated with the setpoints used for ESF actuation functions.

The setpoint methodology follows the methodology in ANSI/ISA-67.04-1994 (Reference 27). 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(6)).

12.3.4.1.7 Calibration Methods and Frequency The methodology to determine the calibration methods and frequency of the ARMS is provided by the ODCM based on plant procedures.

12.3.4.1.8 Power Supplies Instrument loops of safety-related monitors are powered from the appropriate train of Class 1E 120 AC distribution panel in the instrument power system (IP), which is powered by the onsite Class 1E emergency diesel generator. When the emergency diesel generator restores power to the skid, skid equipment such as sample pumps returns to the original operating status without having to be manually restarted. The TSC area radiation monitor, which is non-safety-related, is powered from permanent non-safety buses that are backed up by an alternate ac generator. Instrumentation and control power are described further in Subsection 8.3.2.

12.3.4.2 Airborne Radioactivity Monitoring Instrumentation Airborne radioactivity monitors are installed in selected areas and HVAC systems to provide plant operating personnel with continuous information on the airborne radioactivity levels throughout the plant. These monitors, consisting of gaseous process and effluent radiation monitors (PERMS), are described in Section 11.5 and listed in Table 11.5-1. The airborne radioactivity monitors are as follows:

a. High-energy line break area HVAC exhaust monitor

b. Auxiliary building controlled area common HVAC exhaust monitor 12.3-41 Rev. 2

APR1400 DCD TIER 2

c. Containment purge exhaust monitor

d. Containment air monitor

e. Fuel handling area HVAC exhaust monitor

f. Condenser vacuum vent exhaust monitor

g. Main control room air intake monitor

h. Compound building HVAC exhaust monitor

i. Miscellaneous process monitors 12.3.4.2.1 Design Objectives The objectives of the airborne radioactivity monitors are presented in Subsection 11.5.1.1.

12.3.4.2.2 Location of Airborne Radioactivity Monitors The criteria for the location of the airborne radioactivity monitors are presented in Subsection 11.5.1.1 and the monitor locations are shown in Figure 11.5-2.

12.3.4.2.3 System Description Airborne radioactivity monitors and applicable design criteria are described in Subsection 11.5.1.2.

12.3.5 Dose Assessment The dose assessment is described in Section 12.4.

12.3.6 Combined License Information 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.

12.3-42 Rev. 2

APR1400 DCD TIER 2 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

- Transfer tube access area manway hatch in Room (137-A40B) at elevation 137'-6"

- Doors between Truck Bay (100-P08) and the building exterior Also, the COL applicant is to provide an ITAAC for the radiation shielding and 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.

12.3-43 Rev. 2

APR1400 DCD TIER 2 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.

12.3.7 References

1. Regulatory Guide 8.8, Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations will be ALARA, Rev. 3, U.S. Nuclear Regulatory Commission, June 1978.

2. Regulatory Guide 8.10, Operating Philosophy for Maintaining Occupational Radiation Exposures As Low As Is Reasonably Achievable, Rev. 1-R, U.S. Nuclear Regulatory Commission, May 1977.

3. 10 CFR Part 20, Standards for Protection against Radiation, U.S. Nuclear Regulatory Commission.

4. Regulatory Guide 4.21, Minimization of Contamination and Radioactive Waste Generation: Life-Cycle Planning, U.S. Nuclear Regulatory Commission, June 2008.

5. Regulatory Guide 8.25, Air Sampling in the Workplace, Rev. 1, U.S. Nuclear Regulatory Commission, June 1992.

6. 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, U.S.

Nuclear Regulatory Commission.

7. NUREG-0737, Clarification of TMI Action Plan Requirements U.S. Nuclear Regulatory Commission.

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

9. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluation Design Basis Accidents at Nuclear Power Reactors, U.S. Nuclear Regulatory Commission, July 2000.

12.3-44 Rev. 2

APR1400 DCD TIER 2

10. Regulatory Guide 1.69, Concrete Radiation Shields and Generic Shield Testing for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, Rev. 1, May 2009.

11. 10 CFR 50.34, Contents of Applications, Technical Information, U.S. Nuclear Regulatory Commission.

12. 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.

13. ANISN, A One Dimensional Discrete-Ordinates Transport Code with Anisotropic Scattering, Union Carbide Corporation, 1967.

14. RSICC Computer Code Collection CCC-740 MCNPX, Monte Carlo N-Particle Transport Code System, Los Alamos National Laboratory, 2006.

15. Grove Software, Inc., MICROSHIELD Users Manual Version 9.05, 2011.

16. RUNT-G, A PC Version of RUNT-G, KEPCO E&C, 1988.

17. ISOSHLD, A Computer Code for General Purpose Isotope Shielding Analysis, BNWL-236, Pacific Northwest Laboratory, Richmond, Washington, 1966.

18. Regulatory Guide 8.38, Control of Access to High and Very High Radiation Areas in Nuclear Power Plants, Rev. 1, U.S. Nuclear Regulatory Commission, May 2006.

19. 10 CFR Part 70, Domestic Licensing of Special Nuclear Material, U.S. Nuclear Regulatory Commission.

20. Regulatory Guide 1.21, Measuring, Evaluating, and Reporting Radioactive Material in Liquid and Gaseous Effluents and Solid Waste, Rev. 2, U.S. Nuclear Regulatory Commission, June 2009.

21. Regulatory Guide 8.2, Guide for Administrative Practices in Radiation Monitoring Measuring, Rev. 1, U.S. Nuclear Regulatory Commission, May 2011.

22. ANSI/HPS N13.1, Sampling and Monitoring Releases of Airborne Radioactive Substance From the Stack and Ducts of Nuclear Facilities, Health Physics Society, January 1999.

12.3-45 Rev. 2

APR1400 DCD TIER 2

23. IEEE Std. 497-2002, Accident Monitoring Instrumentation for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 2002.

24. ANSI/ANS HPSSC 6.8.1, Location and Design Criteria for Area Radiation Monitoring Systems for Light Water Nuclear Reactors, American Nuclear Society, May 1981.

25. 10 CFR 50.68, Criticality Accident Requirements, U.S. Nuclear Regulatory Commission.

26. Regulatory Guide 1.206, Combined License Applications for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, June 2007.

27. ANSI/ISA-67.04-1994, Setpoints for Nuclear Safety-Related Instrumentation, International Society of Automation, 1994.

28. IEEE Std. 323-2003, IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 2003.

29. IEEE Std. 344-2004, IEEE Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 2004.

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

31. 10 CFR 52.47, Contents of Applications; Technical Information, U.S. Nuclear Regulatory Commission.

12.3-46 Rev. 2

APR1400 DCD TIER 2 Table 12.3-1 Cobalt Contents of Finished Surface for Primary System Materials Component Cobalt Content Reactor Vessel Assembly Cladding(1) 0.05 w/o (max)

Reactor Coolant Pipe Cladding(1) 0.05 w/o (max)

(1)

Pressurizer Assembly Cladding 0.05 w/o (max)

(1)

Steam Generator Cladding 0.05 w/o (max)

(1)

Reactor Coolant Pump Cladding 0.05 w/o (max)

(6)

Reactor Coolant Pump Impeller 0.05 w/o (max)

(2)

Core Support Barrel Assembly 0.05 w/o (max)

(2)

Lower Support Structure Assembly 0.05 w/o (max)

(2)

Instrument Nozzle Assembly 0.05 w/o (max)

(2)

Upper Guide Structure Assembly 0.05 w/o (max)

(2)

Inner Barrel Assembly 0.05 w/o (max)

(2)

Core Shroud Assembly 0.05 w/o (max)

(4)

Surge Line 0.05 w/o (max)

Control Element Driving Mechanisms 0.05 w/o (max) including RV CEDM nozzles(2),(3)

Steam Generator Tube(3) 0.015 w/o (mean value)

The components except where identified in Not Limited this table (1) Austenitic Stainless Steel Type 308, or Type 309, or Ni based Alloy (Alloy690 equivalent weld material)

(2) Austenitic Stainless Steel Type 304 (3) Alloy 690TT (4) Austenitic Stainless Steel Type 347 (5) Small quantity of cobalt base alloy (Stellite or Haynes alloy) as bar, casting, or hardfacing is used for the CEDMs, RVIs, pumps, or valves.

(6) Martensitic Stainless Steel F6MN or equivalent 12.3-47 Rev. 2

APR1400 DCD TIER 2 Table 12.3-1a Cobalt Contents of Fuel Assembly Components Materials Components Cobalt Content Top Nozzle except Holddown Spring(1) 0.05 w/o (max)

Top Nozzle Holddown Spring(2) 0.10 w/o (max)

(1)

Bottom Nozzle 0.05 w/o (max)

(2)

Top, Bottom and Protective Grids 0.04 w/o (max)

Top and Bottom Grid Sleeves(1) 0.12 w/o (max)

Guide Thimble Screw(1) 0.12 w/o (max)

(1) Austenitic Stainless Steel Type 304/304L (2) Nickel Alloy 718 (3) Small surface components compared to Fuel Rod 12.3-48 Rev. 2

APR1400 DCD TIER 2 Table 12.3-2 Normal Operation Radiation Zone Designations Zone Design Dose Rate Designation(1) (mSv/hr)(2) Zone Description 1 DR 0.0025 Uncontrolled, unlimited access 2 0.0025 < DR 0.025 Controlled, limited access, 40 hr/wk 3 0.025 < DR 0.05 Controlled, limited access, 20 ~ 40 hr/wk 4 0.05 < DR 0.2 Controlled, limited access, 5 ~ 20 hr/wk 5 0.2 < DR 1 Controlled, limited access, 1 ~ 5 hr/wk 6 1 < DR 10 Controlled access under supervision of radiation protection personnel 7 10 < DR 5,000 Controlled access under supervision of radiation protection personnel 8 DR > 5,000 Controlled access under supervision of radiation protection personnel (1) High-radiation areas include radiation Zones 6, 7, and 8.

(2) DR denotes dose rate at 30 cm (11.8 in.) from the radiation source or from any surface that the radiation penetrates.

12.3-49 Rev. 2

APR1400 DCD TIER 2 Table 12.3-3 Post-Accident Radiation Zone Designations and Design Dose Rates Zone Design Dose Rate Designation (mSv/hr) 1 DR 0.15 2 0.15 < DR 1 3 1 < DR 10 4 10 < DR 100 5 100 < DR 1,000 6 1,000 < DR 5,000 7 DR > 5,000 12.3-50 Rev. 2

APR1400 DCD TIER 2 Table 12.3-4 (1 of 16)

Design Basis Radiation Shield Thicknesses Room Minimum Required Shield Thickness (inches)

Number Room Name North South East West Floor Ceiling Containment Building 069-C01 ICI Cavity - 60 79 60 Ground -

078-C01 Reactor Cavity 80 80 79 79 Ground -

080-C01 Holdup Volume Tank 36 36 - 79 Ground -

100-C02 Steam Generator Cavity 36 36 30 36 Ground -

100-C03 Reactor Drain Tank 24 24 48 24 - 24 100-C04 Letdown HX Room 24 24 48 24 - 24 128-C01 Regenerative HX Room 60 60 48 48 24 48 136-C02 Pressurizer Cavity 33 36 30 33 30 -

Auxiliary Building 050-A01C CS Pump & Miniflow HX Room 48 48 48 24 Ground 30 050-A01D CS Pump & Miniflow HX Room 48 48 48 24 Ground 30 050-A02C SI Pump Room 48 48 48 48 Ground 30 050-A02D SI Pump Room 48 48 48 48 Ground 30 050-A03A SI Pump Room 48 48 24 48 Ground 30 050-A03B SI Pump Room 48 48 24 48 Ground 30 050-A04A SC Pump and Miniflow HX Room 48 48 27 24 Ground 30 050-A04B SC Pump and Miniflow HX Room 48 48 27 24 Ground 30 055-A01C CS HX Room 36 48 48 30 Ground 30 055-A01D CS HX Room 48 36 48 30 Ground 30 055-A08C Floor Drain Sump Pump Room 14 10 14 10 Ground 14 055-A08D Floor Drain Sump Pump Room 10 14 14 10 Ground 14 055-A10C Tendon Gallery Entrance 24 36 48 48 Ground 24 055-A14C Pipe Chase and Valve Room 48 48 48 48 Ground 30 055-A14D Pipe Chase and Valve Room 48 48 48 48 Ground 30 055-A18A Pipe Chase and Valve Room 42 48 48 20 Ground 30 055-A18B Pipe chase and Valve Room 48 42 48 10 Ground 30 055-A21A Pipe Chase and Valve Room 10 48 25 29 Ground 32 055-A21B Pipe Chase and Valve Room 48 10 20 10 Ground 32 12.3-51 Rev. 2

APR1400 DCD TIER 2 Table 12.3-4 (2 of 16)

Room Minimum Required Shield Thickness (inches)

Number Room Name North South East West Floor Ceiling Auxiliary Building (cont.)

055-A22A Pipe Chase 10 10 33 42 Ground 10 055-A22B Pipe Chase 10 10 48 42 Ground 10 055-A30A SC HX Room 36 48 48 48 Ground 30 055-A30B SC HX Room 48 36 48 48 Ground 30 055-A31B Chemical Drain Sump Pump Room 27 10 10 10 Ground 20 055-A33A Equipment Drain Sump Pump Room 10 22 14 10 Ground 14 055-A33B Equipment Drain Sump Pump Room 19 10 10 10 Ground 17 055-A34A Floor Drain Sump Pump Room 10 22 10 22 Ground 11 055-A34B Floor Drain Sump Pump Room 22 10 10 22 Ground 18 055-A38A Boronometer Room 25 25 34 25 Ground 28 055-A39A Process Radiation Monitor Room 19 19 11 19 Ground 21 055-A42A Charging Pump Room 17 24 24 28 Ground 22 055-A43A Charging Pump Miniflow HX Room 10 18 10 10 Ground 19 055-A45A Pipe Chase 27 17 25 10 Ground 24 055-A46A Condensate Return Unit Room 28 37 10 16 Ground 16 055-A47B Primary Off-Gas Sample Pump Room 72 48 60 24 Ground 24 055-A49B Post Accident Sample Control Panel 24 72 18 60 Ground 24 RM 055-A51B Equipment Drain Tank Room 18 36 48 36 Ground 24 055-A52B Reactor Drain Pump Room 18 18 18 36 Ground 24 055-A53B Reactor Drain Pump Room 13 12 23 10 Ground 17 055-A54B Aux. Charging Pump Room 66 24 60 24 Ground 24 055-A55B Charging Pump Room 65 66 60 60 Ground 24 055-A56A Valve Room 10 18 28 10 Ground 19 055-A56B Valve Room 14 20 12 10 Ground 10 055-A58A Pipe Chase 18 10 26 26 Ground 10 055-A59A Valve Room 19 10 26 34 Ground 24 055-A62A Chase 18 48 18 27 Ground 30 055-A62B Chase 48 18 18 27 Ground 30 068-A06A Gas Stripper Room 37 35 45 35 30 24 12.3-52 Rev. 2

APR1400 DCD TIER 2 Table 12.3-4 (3 of 16)

Room Minimum Required Shield Thickness (inches)

Number Room Name North South East West Floor Ceiling Auxiliary Building (cont.)

068-A07A Hot Pipe Way 13 29 31 29(1)/ 24 24 24(2)/

21(3) 068-A08B Hot Pipe Way 33 24 24 20 23 30 068-A09B Valve Room 19 17 23 19 17 10 068-A10A Filter and Demin. Valve Area 10 10 33 33 30 24 068-A11A Filter and Demin. Valve Area 10 24 10 22 24 21 068-A12A Filter and Demin. Valve Area 10 18 10 18 17 29 077-A01A Reactor Drain Filter Pit 35 10 10 15 10 23 077-A02A SFP Cleanup Filter Pit 10 10 10 10 23 23 077-A03A SFP Demin Filter Pit 10 17 10 18 23 23 077-A04A SFP Cleanup Filter Pit 10 10 10 10 23 23 077-A05A SFP Demin. Filter Pit 10 10 10 10 23 23 077-A06A Purification Filter Pit 10 10 10 10 10 39 077-A07A Reactor Makeup Water Filter Pit 18 32 18 18 10 10 077-A08A Purification Filter Pit 10 10 10 10 10 39 077-A09A SGBD Filter Pit 10 10 10 10 22 23 077-A10A Seal Injection Filter Pit 10 10 10 11 24 24 077-A11A SGBD Filter Pit 10 10 10 10 22 23 077-A12A Seal Injection Filter Pit 10 13 10 10 13 24 077-A13A SGBD Filter Pit 35 10 10 10 22 23 077-A14A Boric Acid Filter Pit 13 18 10 29 10 22 077-A15A Filter Cartridge Storage 32 39 39 39 25 39 078-A09C HAVC Chase EL. 100-0 18 30 48 48 18 18 EL. 156-0 18 24 48 36 18 18 078-A09D HAVC Chase EL. 100-0 30 18 48 48 18 18 EL. 137-6 18 18 48 48 18 18 EL. 156-0 24 18 48 36 18 18 (1) West wall within the row line from AD to AF along the column line 24 (2) Wall thickness for labyrinth design (3) West wall within the column line from 25 to 26 and row line from AA to AD 12.3-53 Rev. 2

APR1400 DCD TIER 2 Table 12.3-4 (4 of 16)

Room Minimum Required Shield Thickness (inches)

Number Room Name North South East West Floor Ceiling Auxiliary Building (cont.)

078-A12C Essential Water Chiller RM 36 48 48 48 30 24 078-A12D Essential Water Chiller RM 48 36 48 48 30 24 078-A13D Duct RM 36 24 48 48 18 24 078-A14C Buttress Opng EL. 100-0 18 18 48 48 18 -

EL. 137-6 18 18 48 48 18 -

EL. 156-0 18 18 48 36 18 -

078-A15C Turbine Driven Aux. Feedwater Pump 48 48 48 48 30 48 RM 078-A15D Turbine Driven Aux. Feedwater Pump 48 48 48 48 30 48 RM 078-A20A Motor Driven Aux Feedwater Pump 48 48 48 48 30 48 RM 078-A20B Motor Driven Aux Feedwater Pump 48 48 48 48 30 48 RM 078-A21A Pipe Chase 10 36 48 48 10 24 078-A21B Pipe Chase 36 48 48 48 24 24 078-A25A Class 1E Switchgear 01A RM 36 48 48 30 30 30 078-A25B Class 1E Switchgear 01B RM 48 36 48 30 30 30 078-A32A SPF Cleanup Demin. Room 10 22 22 21 12 24 078-A33A SG Blowdown Demin. Room 33 10 23 10 10 23 078-A34A Pre-Holdup IX Room 33 15 15 33 15 24 078-A35A Purification IX Room 12 24 29 43 29 42 078-A36A Boric Acid IX Room 24 18 18 18 18 18 078-A37A Deborating IX Room 15 12 10 24 24 24 078-A38A SFP Cleanup Pump Room 23 23 23 10 23 32 078-A39A Gas Stripper Effluent Radiation 23 23 23 10 16 35 Monitor Room 078-A40B Boric Acid Concentrator Room 16 23 23 14 16 16 086-A01A Filter Area - - 18 21 13 10 100-A32B SFP Cooling HX Room 10 10 10 10 10 10 100-A29B Pipe and HVAC Chase 10 12 10 66 10 10 100-A07C AUX. Feedwater Tank 48 48 48 48 48 48 100-A07D AUX. Feedwater Tank 48 48 48 48 48 48 12.3-54 Rev. 2

APR1400 DCD TIER 2 Table 12.3-4 (5 of 16)

Room Minimum Required Shield Thickness (inches)

Number Room Name North South East West Floor Ceiling Auxiliary Building (cont.)

100-A08C N1E DC & IP Equipment RM 36 48 48 48 48 18 100-A08D N1E DC & IP Equipment RM 48 30 48 48 24 18 100-A09C Tendon Access RM 30 36 48 48 24 18 100-A13A Mechanical Penetration Room 36 48 48 48 30 18 100-A13B Mechanical Penetration Room 48 36 48 48 30 18 100-A14A Personnel Air Lock Enterance 48 36 48 48 48 18 100-A16D Pipe Chase 48 48 48 48 10 23 100-A16C Pipe Chase 48 48 48 48 13 10 100-A24A SFP Cooling HX Room 12 10 12 40 24 10 100-A26A Valve Room 28 41 21 28 32 10 100-A25A Volume Control Tank Room 42 42 42 47 48 53 111-A01B Cask Loading Pit 48 14 48 48 42 -

114-A01B Spent Fuel Pool 62 60 59 68 71 -

113-A01B Fuel Transfer Tube Inspection Area 44 44 No walls 62 60 (1) 119-A01B Refueling Canal 60/30 59 62 48 62 -

120-A09C Electrical Penetration RM (C) 18 48 48 48 18 18 120-A16B Mechanical Penetration Room 29 27 33 48 18 29 120-A16A Mechanical Penetration Room 20 24 20 48 17 19 120-A23A Valve Room 18 25 18 18 10 18 120-A14A SG Blowdown Regen. HX Room 48 48 48 48/ 48 18 27 (2) 120-A21A Aux. Bldg Controlled Area (I) 18 48 - 36 18 18 Emerge ncy Exhaust ACU RM-1 120-A32A Aux. Bldg Controlled Area (I) 18 48 48 - 18 18 Emerge ncy Exhaust ACU RM-2 137-A19A SG Blowdown Flash Tank Room 42 42 48 48 48 18 137-A11D Electrical Penetration RM (D) 48 18 48 48 18 18 137-A31C Main Steam Valve EL. 137-6 48 60 48 48/ 48 18 RM 84 (3)

EL. 156-0 48 60 48 48 48 18 (1) Minimum required shield thicknesses of the portions in contact with spent fuel pool (114-A01B) and pipe

& HVAC chase (100-A29B) are 60 inches and 30 inches, respectively.

(2) Shield thickness of the portions in contact with general access area (120-A20A) & Stair (055-A20A) are 48 inches and 27 inches, respectively.

(3) Shield thickness of the portions in contact with 480V Class 1E MCC 03C room (137-A10C) & main steam enclosure (137-A30C) are 48 inches and 84 inches, respectively.

12.3-55 Rev. 2

APR1400 DCD TIER 2 Table 12.3-4 (6 of 16)

Room Minimum Required Shield Thickness (inches)

Number Room Name North South East West Floor Ceiling Auxiliary Building (cont.)

137-A31D Main Steam Valve RM (EL. 156-0) 60 48 48 48 48 18 156-A14A Aux. Bldg Controlled Area (I) Normal 22 22 22 22 22 22 Exhaust ACU Room 157-A19C I&C Equip. RM 18 36 48 36 18 18 157-A19D I&C Equip. RM 36 18 48 36 18 18 157-A20C I&C Equip. RM 24 18 48 36 18 18 157-A20D I&C Equip. RM 18 24 48 36 18 18 174-A15B Containment High- and Low-volume 21 21 21 21 15 10 Purge ACU Room 195-A08B Aux. Bldg. Controlled Area (II) 22 22 22 22 18 22 Normal Exhaust ACU Room 12.3-56 Rev. 2

APR1400 DCD TIER 2 Table 12.3-4 (7 of 16)

Room Minimum Required Shield Thickness (inches)

Number Room Name North South East West Floor Ceiling Compound Building 063-P01 Hot Pipe Chase 10 28 28 10 Ground 17 063-P02 GRS Header Drain Tank Room 35 42 48 41 Ground 18 063-P03 Valve Room 27 30 37 10 Ground 10 063-P04 GRS Inlet Skid Room 22 34 10 22 Ground 13 063-P05 Spent Resin Long-term Storage Tank 27 35 48 36 Ground 46 Room 063-P06 Future Use 36 27 48 36 Ground 39 063-P07 Valve Room 16 29 36 30 Ground 14 063-P08 Low-activity Spent Resin Tank Room 27 32 35 10 Ground 10 063-P09 Valve Room 16 36 10 16 Ground 18 063-P13 Hot Pipe Chase 40 33 40 33 Ground 19 063-P14 Hot Tool Room 15 10 10 10 Ground 32 063-P21 Equip. Waste Pump Room 17 19 10 20 Ground 17 063-P22 Equip. Waste Pump Room 10 17 10 21 Ground 17 063-P23 Equip. Waste Tank Room 13 33 20 22 Ground 27 063-P24 Equip. Waste Tank Room 16 13 21 22 Ground 27 063-P25 Floor Drain Pump Room 14 10 11 19 Ground 10 063-P26 Normal Sump Pump Room 14 14 10 19 Ground 16 063-P27 Chemical Waste Pump Room 10 14 10 15 Ground 16 063-P28 Floor Drain Tank Room 16 16 19 20 Ground 29 063-P29 Floor Drain Tank Room 16 16 19 20 Ground 29 063-P30 Chemical Waste Tank Room 10 16 15 10 Ground 15 063-P31 Chemical Waste Tank Room 10 10 15 10 Ground 15 063-P36 DWS Drain Sump Pump Room 10 10 10 10 Ground 10 12.3-57 Rev. 2

APR1400 DCD TIER 2 Table 12.3-4 (8 of 16)

Room Minimum Required Shield Thickness (inches)

Number Room Name North South East West Floor Ceiling Compound Building (cont.)

063-P37 Monitor Tank Room 10 18 11 18 Ground 10 063-P38 PSS-Solidification & Drum Conveyer 17 24 24 21 Ground 14 Room 063-P39 Spent Resin Long-term Storage Tank 18 20 18 21 Ground 18 Sump Pump Room 063-P40 Concentrate Pump Room 27 24 20 16 Ground 19 063-P41 Concentrate Holding Tank Room 21 27 33 28 Ground 10 063-P42 RO Feed Pump Room 10 10 28 16 Ground 24 063-P43 IX Feed Pump Room 16 10 16 10 Ground 24 063-P44 IX Feed Tank Room 14 16 11 10 Ground 23 063-P47 CTS HEPA Vacuum Skid Room 24 10 21 10 Ground 10 063-P48 CTS Dryer Skid Room 31 24 17 21 Ground 15 063-P49 CTS Vacuum Skid Room 10 10 21 10 Ground 18 063-P54 Monitor Tank Pump Room 10 10 10 10 Ground 14 063-P73 Future Use 36 43 18 48 Ground 36 085-P01 Waste Gas Dryer Skid Room 18 31 31 25 17 24 085-P02 Waste Gas Dryer Skid Room 11 18 22 25 17 24 085-P03 Valve Room 48 11 30 26 18 36 085-P04 Charcoal Guard Bed Room 27 27 35 18 10 24 085-P06 Valve Room 19 26 36 30 19 27 085-P07 Valve Room 27 24 30 30 14 24 085-P08 Valve Room 24 19 22 24 19 24 085-P15 Valve Room 10 21 21 18 10 23 085-P16 Valve Room 10 22 18 19 10 22 085-P17 Valve Room 10 10 10 14 10 17 085-P20 Valve Room 16 16 16 16 10 16 085-P21 Charcoal Guard Bed Room 27 27 18 35 10 24 085-P31 Primary Sampling Room 10 10 10 10 10 10 12.3-58 Rev. 2

APR1400 DCD TIER 2 Table 12.3-4 (9 of 16)

Room Minimum Required Shield Thickness (inches)

Number Room Name North South East West Floor Ceiling Compound Building (cont.)

085-P32 Primary Sampling Sink Room 10 13 12 12 14 18 085-P42 IX Module Room 10 30 30 27 14 28 085-P43 IX Module Room 30 10 30 30 14 28 085-P44 RO Feed Tank Room 10 27 32 22 19 25 085-P45 Drum Removal Chase 15 15 15 15 - 25 085-P46 MF Membrane Module Room 23 10 20 15 18 16 085-P47 MF Membrane Module Room 23 16 10 12 15 16 085-P48 RO Membrane Module and Valve 43 24 43 34 32 36 Skid Room 096-P01 Charcoal Delay Bed Room 49 47 49 15 36 43 096-P02 Charcoal Delay Bed Room 49 47 15 40 36 43 100-P02 GRS Equipment Removal Area 13 11 40 10 23 10 100-P07 Future Extension Area 24 30 36 37 24 31 100-P08 Truck Bay 24 24 36 37 36 31 100-P09 Waste Drum Storage Area 28 24 36 26 34 31 100-P10 Spent Filter Drum Storage Area 36 28 48 37 36 43 120-P01 Gaseous Radwaste Sample Control 16 21 21 12 18 26 Panel Room 120-P02 Gaseous Radwaste Sample Valve 22 16 26 18 18 26 Rack Room 139-P06 Normal Exhaust ACU Room 20 20 20 20 20 20 Yard

- Boric Acid Storage Tank 16 16 16 16 - -

(1) (1) (1) (1)

- Holdup Tank 15 15 15 15 - -

(1) Including the Tank wall of 0.25 inches 12.3-59 Rev. 2

APR1400 DCD TIER 2 Table 12.3-4 (10 of 16)

Minimum Face to Required Shield Room Room Room Thickness Number Name Structure Number Room Name (inches)

Additional Information on the Rooms which Contain Complex Geometries 068- Hot Slab 055-A45A Pipe Chase 10 A07A Pipe Way Slab 055-A42A Charging Pump Room 22 Slab 055-A35A General Access Area 24 Slab 055-A39A Process Radiation Monitor Room 21 Slab 055-A58A Pipe Chase 17 Slab 055-A30A SC Hx Room 19 Slab 078-A38A Spent Fuel Pool Clean up Pump 23 Room Slab 078-A39A Gas Stripper Effluents Radiation 16 Monitor Room Slab 078-A31A General Access Area 24 068- Filter Slab 055-A35A General Access Area 30 A10A and Slab 055-A36A CVCS Chemical Package Room 30 Demin Valve Slab 078-A34A Pre-Holdup Ion Exchange 15 Room Slab 078-A35A Purification Ion Exchange Room 13 Slab 078-A37A Deborating Ion Exchange Room 24 Slab 077-A01A Reactor Drain Filter Pit 15 Slab 077-A06A Purification Filter Pit 10 Slab 077-A08A Purification Filter Pit 10 Slab 077-A10A Seal Injection Filter Pit 10 12.3-60 Rev. 2

APR1400 DCD TIER 2 Table 12.3-4 (11 of 16)

Face to (floor) Minimum Required Shield Room Room Room Thickness Number Name Structure Number Room Name (inches)

Additional Information on the Rooms which Contain Complex Geometries (cont.)

077-P01 Hot Slab 063-P27 Chemical Waste Pump Room 14 Pipe Slab 063-P33 Sampling Room 14 Way Slab 063-P34 LRS Control Panel Room 14 Slab 063-P54 Monitor Tank Pump Room 14 Slab 063-P56 Drop Area & Opening 14 Slab 063-P57 Sorting Room 14 Slab 063-P21 Equip. Waste Pump Room 15 Slab 063-P22 Equip. Waste Pump Room 17 Slab 063-P25 Floor Drain Pump Room 10 Slab 063-P26 Normal Sump Pump Room 16 Slab 063-P38 PSS-solidification & Drum 16 Conveyor Room Slab 063-P46 Corridor 25 Slab 063-P11 Corridor 28 Slab 063-P09 Valve Room 20 Slab 063-P40 Concentrate pump Room 20 Slab 063-P41 Concentrate holding tank Room 10 Slab 063-P42 RO feed pump Room 24 Slab 063-P43 IX feed pump Room 24 Slab 063-P44 IX Feed tank Room 23 Slab 063-P45 Hot tool Room 23 Slab 063-P04 GRS inlet skid Room 13 Slab 063-P07 Valve Room 14 Slab 063-P10 Hot tool Room 24 (1)

Slab 063-P16 Corridor 18 (2)

Slab 063-P16 Corridor 24 (1) Section within the column line from 33 to 36 and from PF to PG (2) Section within the column line from 36 to 37 and from PB to PG 12.3-61 Rev. 2

APR1400 DCD TIER 2 Table 12.3-4 (12 of 16)

Minimum Required Shield Openings for Thickness (inches)

Room Room Room Number Name Structure Number Room Name North South East West Additional Information on the Rooms which Contain Complex Geometries (cont.)

077-P01 Hot Wall 063-P26 Normal Sump 16 16 16 19 Pipe Room Way Wall 063-P56 Drop Area 14 14 18 14 Wall 085-P45 Drum 16 16 16 16 Removal Chase Wall 063-P16 Corridor(1) 24 - 24 20 Opening within the column line from 36 to 37 and PE to PG Wall 063-P16 Corridor 18 24 - 24 Wall 057-P01 Elev. Hoist 18 - 34 -

Way Wall 063-P18 Stair - 32 32 -

Opening within the column line from 36 to 37 and PB to PE Wall 063-P16 Corridor 20 20 - 20 Wall 063-P19 Elect. Riser 29 - 29 -

Wall 063-P39 Spent Resin 18 21 15 21 Long Term Storage Pump Room Wall 063-P20 HVAC Chase - 18 30 -

(1) Opening within the column line from 36 to 37 and from PA to PB 12.3-62 Rev. 2

APR1400 DCD TIER 2 Table 12.3-4 (13 of 16)

Face to (side wall) Minimum Required Shield Room Room Room Thickness Number Name Structure Number Room Name (inches)

Additional Information on the Rooms which Contain Complex Geometries (cont.)

077-P01 Hot Wall - Yard(1) 28 Pipe Wall 063-P23 Equipment Waste Tank Room 15 Way Wall 063-P24 Equipment Waste Tank Room 15 Wall 063-P28 Floor Drain Tank Room 16 Wall 063-P29 Floor Drain Tank Room 19 Wall 063-P30 Chemical Waste Tank Room 10 Wall 063-P31 Chemical Waste Tank Room 10 (2)

Wall 063-P16 Corridor 18 Wall 063-P52 Chemical Drain Sump Pump 14 Room Wall 062-P02 Mask Decontamination Room 14 Wall 063-P64 Corridor 14 Wall 063-P61 Laundry Storage 14 Wall 063-P51 Stair 17 Wall 063-P78 Pipe Chase 14 Wall 063-P32 Detergent Waste Tank & Pump 14 Room Wall 063-P37 Monitor Tank Room 18 (3)

Wall 063-P16 Corridor 18 Wall 063-P46 Corridor

- Wall within the column line 18 from 37 to 38 along the row line PG

- Wall within the row line from 25 PG to PH along the column line 38 Wall 063-P38 PSS-Solidification 16 Wall 085-P45 Opening for Drum Removal 16 Chase (West wall)

Wall 063-P48 CTS-Dryer Skid Room 17 Wall 063-P13 Hot Pipe Chase 10 (1) Exterior wall within the column line from 35 to 36 along the row line PA (2) Section within the column line from 33 to 35 and from PF to PG (3) Side wall within the column line from 36 to 37 along the row line PG 12.3-63 Rev. 2

APR1400 DCD TIER 2 Table 12.3-4 (14 of 16)

Face to (side wall) Minimum Required Shield Room Room Room Thickness Number Name Structure Number Room Name (inches)

Additional Information on the Rooms which Contain Complex Geometries (cont.)

077-P01 Hot Wall - Yard(4) 38 Pipe Wall 063-P11 Corridor (5)

Way

- South wall 28

- East wall 28 Wall 063-P08 Low Activity Spent Resin Tank 10 Room Wall 063-P06 Future Use Area 32 Wall 063-P05 Spent Resin Long term Storage 32 Tank Room Wall 063-P02 GRS Header Drain Tank 24 (6)

Wall 063-P16 Corridor 24 Wall 063-P01 Hot Pipe Chase 17 (7)

Wall - Yard 32 (4) Exterior wall within the row line from PF to PI along the column line 39 (5) Section within the column line from 38 to 39 and from PE to PF (6) Section within the column line from 38 to 39 and from PB to PC (7) Exterior wall within the column line from 37 to 38 along the row line PA 12.3-64 Rev. 2

APR1400 DCD TIER 2 Table 12.3-4 (15 of 16)

Face to (side wall) Minimum Required Shield Room Room Room Thickness Number Name Structure Number Room Name (inches)

Additional Information on the Rooms which Contain Complex Geometries (cont.)

077-P01 Hot Slab 085-P14 Corridor(1) 18 Pipe Slab 085-P17 Valve Room 10 Way Slab 085-P20 Valve Room 10 Slab 085-P31 Primary Sampling Room 10 Slab 085-P32 Primary Sampling Sink Room 10 Slab 085-P33 Hot Tool Room 16 Slab 085-P35 Storage 14 Slab 085-P15 Valve Room 10 Slab 085-P16 Valve Room 10 Slab 085-P44 RO Feed Tank Room 21 Slab 085-P45 Drum Removal Chase 16 Slab 085-P46 MF Membrane Module Room 20 Slab 085-P07 Valve Room 16 Slab 085-P08 Valve Room 22 (2)

Slab 085-P14 Corridor 30 (3)

Slab 085-P14 Corridor 24 Slab 085-P42 IX Module Room 16 Slab 085-P43 IX Module Room 16 Slab 085-P01 Waste Gas Dryer Skid Room 15 Slab 085-P02 Waste Gas Dryer Skid Room 16 Slab 085-P03 Valve Room 19 Slab 085-P04 Charcoal Guard Bed Room 10 Slab 085-P06 Valve Room 22 Slab 085-P21 Charcoal Guard Bed Room 10 (1) Section within the column line from 35 to 37 and from PF to PG (2) Section within the column line from 37 to 38 and from PE to PF (3) Section within the column line from 36 to 37 and from PB to PE 12.3-65 Rev. 2

APR1400 DCD TIER 2 Table 12.3-4 (16 of 16)

Minimum Required Shield Piping Area Description Thickness (inches)

Additional Information on the Rooms which Contain Complex Geometries (cont.)

Piping area in the North wall within the column line from 14 to 15 along the row 36 mezzanine floor line AJ adjacent to CSHX North wall within the column line from 15 to 16 along the row 48 RM (055-A01D) line AJ East wall 36 South wall within the column line from 15 to 16 along the row 48 line AI West wall within the row line from AI to AJ along the column 48 line 15 South wall within the column line from 14 to 15 30 West wall within the row line from AI to AJ along the column 42 line 14 Floor 24 Ceiling 30 Piping area in the South wall within the column line from 14 to 15 along the row 36 mezzanine floor line AB adjacent to CSHX South wall within the column line from 15 to 16 along the row 48 RM (055-A01C) line AB East wall 36 North wall within the column line from 15 to 16 along the row 48 line AC West wall within the row line from AB to AC along the column 48 line 15 North wall within the column line from 14 to 15 30 West wall within the row line from AB to AC along the column 42 line 14 Floor 24 Ceiling 30 12.3-66 Rev. 2

APR1400 DCD TIER 2 Table 12.3-5 (1 of 2)

Areas Potentially Greater than 1 Gy/hr(1)

Auxiliary and Containment Building El. 55 ft; See Figure 12.3-1.

HRA /

Area Coordinates VHRA(3) Access Control ICI Cavity AF-AG, 18-19 VHRA Locked Door Auxiliary and Containment Building El. 78 ft; See Figures 12.3-2 and 3.

Area Coordinates HRA / VHRA Access Control Hold-Up Volume Tank AE-AG, 20-21 HRA Locked Door Core Debris Chamber VH AE-AG, 17-18 VHRA Locked Door Pre-Holdup Ion Exchanger Pit AC-AD, 23-24 HRA Hatch Purification Ion Exchanger Pit VH AB-AC, 23-24 VHRA Hatch Purification Filter Pit VH AC-AD, 24-25 VHRA Hatch Filter Cartridge Storage VH AA-AB, 24-25 VHRA Hatch Auxiliary and Containment Building El. 100 ft; See Figure 12.3-4.

Area Coordinates HRA / VHRA Access Control Reactor Cavity VH AE-AG, 18-20 VHRA Locked Door Steam Generator Cavity AD-AE & AG-AH, HRA Locked Door Reactor Drain Tank Room 19 HRA Locked Door Volume Control Tank Room VH AE-AF, 16-17 VHRA Locked Door AD-AE, 24-25 Auxiliary and Reactor Containment Building El. 120.0 ft; See Figure 12.3-5.

Area Coordinates HRA / VHRA Access Control (2)

Refueling Pool Area AE-AG, 17-21 VHRA No Entrance Transfer Tube Inspection Area(2) AF-AG, 21-23 VHRA Locked Door Spent Fuel Pool(2) AG-AH, 23-25 VHRA No Entrance Cask Loading Pit(2) AH-AI, 23-24 VHRA No Entrance Refueling Canal(2) AF-AG, 23-25 VHRA No Entrance (1) During normal operating conditions and AOOs (2) Only when fuel is in the area (3) HRA : High Radiation Area, VHRA : Very High Radiation Area (Greater than 5 Gy/hr) 12.3-67 Rev. 2

APR1400 DCD TIER 2 Table 12.3-5 (2 of 2)

Compound Building El. 63 ft; Sdddddee Figure 12.3-10.

Area Coordinates HRA / VHRA Access Control GRS Header Drain Tank Room PB-PC, 38-39 HRA Locked Door Spent Resin Long Term Storage Tank Room PC-PD, 38-39 VHRA Locked Door Future Use PD-PE, 38-39 VHRA Locked Door Hot Pipe Chase PI-PJ, 38-39 HRA Locked Door Future Use(4) PB-PC, 32-33 VHRA(4) Locked Door(4)

Compound Building El. 77 ft; See Figure 12.3-11.

Area Coordinates HRA / VHRA Access Control Hot Pipe Way PA-PI, 33-39 HRA Hatch Compound Building El. 85 ft; See Figure 12.3-12.

Area Coordinates HRA / VHRA Access Control R/O Membrane Module & Valve Skid Room PI-PJ, 37-39 HRA Locked Door Compound Building El. 100 ft; See Figure 12.3-13.

Area Coordinates HRA / VHRA Access Control Charcoal Delay Bed Rooms (096-P01, P02) PB-PC, 37-39 VHRA Hatch Spent Filter Drum Storage Area PI-PJ, 38-39 VHRA Locked Door Truck Bay(5) PF-PG, 37-39 HRA Locked Door Future Extension Area(5) PE-PF, 37-39 HRA Locked Door (4) 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 12.3(7)).

(5) Only during transfer and drumming of spent filter and spent resin 12.3-68 Rev. 2

APR1400 DCD TIER 2 Table 12.3-6 (1 of 2)

Area Radiation Monitors Class(1) Range Function and Remarks Airborne Area(2) Display & Alarm at Description Tag No. S SE E Particulate Iodine Gas Liquid (mSv/hr) MCR/RSR/Local Post-accident RE-205 N III N N/A N/A N/A N/A 10-3~102 Yes/Yes/Yes primary sample room Normal primary RE-285 N III N/A N/A N/A N/A 10-3~102 Yes/Yes/Yes sample room N Main steam and FW RE-237 N II N N/A N/A N/A N/A 100~105 Yes/Yes/Yes containment piping RE-238 penetration area Containment RE-231A 3 I A N/A N/A N/A N/A 10-3 ~ 102 Yes/Yes/Yes operating area RE-232B B CPIAS Containment upper RE-233A 3 I A N/A N/A N/A N/A 101 ~ 108 Yes/Yes/Yes ERDS(3) operating area RE-234B B CPIAS In-core instrument RE-235 N II N N/A N/A N/A N/A 10-3 ~ 102 Yes/Yes/Yes

-3 2 Containment RE-236 N II N N/A N/A N/A N/A 10 ~ 10 Yes/Yes/Yes personnel access hatch area Spent fuel pool area RE-241A 3 I A N/A N/A N/A N/A 10-3 ~ 102 Yes/Yes/Yes RE-242B B FHEVAS New fuel storage RE-245 N II N N/A N/A N/A N/A 10-3 ~ 102 Yes/Yes/Yes area 12.3-69 Rev. 2

APR1400 DCD TIER 2 Table 12.3-6 (2 of 2)

Class(1) Range Function and Remarks Airborne Area(2) Display & Alarm at Description Tag No. S SE E Particulate Iodine Gas Liquid (mSv/hr) MCR/RSR/Local Hot machine shop C-RE- N III N N/A N/A N/A N/A 10-3 ~ 102 Yes/Yes/Yes 293 Radiochemistry lab RE-257 N III N N/A N/A N/A N/A 10-3 ~ 102 Yes/Yes/Yes

-3 2 Main control room RE-275 N II N N/A N/A N/A N/A 10 ~ 10 Yes/Yes/Yes area TSC area RE-279 N III N N/A N/A N/A N/A 10-3 ~ 102 Yes/Yes/Yes Truck bay area C-RE- N III N N/A N/A N/A N/A 10-3 ~ 102 Yes/Yes/Yes 288 C-RE-289 Waste drum storage C-RE- N III N N/A N/A N/A N/A 10-3 ~ 102 Yes/Yes/Yes area 292 Compound building C-RE- N III N N/A N/A N/A N/A 10-3 ~ 102 Yes/Yes/Yes dry active waste 284 storage area (1) S: safety Class per ANSI/ANS-51.1; 1=SC-1, 2=SC-2, 3=SC-3, N=NNS SE: seismic Category I, II, III E: Electrical Class A, B, C, D=Class 1E Separation Division, N=Non-Class 1E Refer to Section 3.2 for the definition.

(2) Detector type for area radiation monitor is GM tube or ionization chamber (3) The monitor transmits the radiation signals to the licensing entity via the emergency response data system (ERDS) link. As described in APR1400 DCD Tier 2 Section 13.3, the ERDS is a real-time electronic data transmission system to the NRC operations center that provides a set of parameters from the onsite computer system in the event of an emergency. The ERDS transmits information to allow the NRC to provide advice and support to the licensee, state, and local authorities, and other federal officials. The ERDS satisfies the requirements in 10 CFR Part 50, Appendix E.

12.3-70 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-1 Radiation Zones(Normal) Auxiliary/Containment Building El.55'-0" 12.3-71 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-2 Radiation Zones (Normal) Auxiliary Building - Partial Plan El. 68'-0", 77'-0", and 86'-0" 12.3-72 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-3 Radiation Zones (Normal) Auxiliary/Containment Building El.78'-0" 12.3-73 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-4 Radiation Zones(Normal) Auxiliary/Containment Building El.100'-0" 12.3-74 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-5 Radiation Zones(Normal) Auxiliary/Containment Building El.120'-0" 12.3-75 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-6 Radiation Zones(Normal) Auxiliary/Containment Building El.137'-6" 12.3-76 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-7 Radiation Zones(Normal) Auxiliary/Containment Building El.156'-0" 12.3-77 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-8 Radiation Zones(Normal) Auxiliary/Containment Building El.174'-0" 12.3-78 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-9 Radiation Zones (Normal) Auxiliary/Containment Building El. 195'-0" and Roof Plan 12.3-79 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-10 Radiation Zones (Normal) Compound Building El. 63'-0" 12.3-80 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-11 Radiation Zones (Normal) Compound Building Partial Plan El. 77-0 12.3-81 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-12 Radiation Zones (Normal) Compound Building El. 85'-0" 12.3-82 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-13 Radiation Zones (Normal) Compound Building El. 100'-0" 12.3-83 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-14 Radiation Zones (Normal) Compound Building El. 120'-0" 12.3-84 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-15 Radiation Zones (Normal) Compound Building El. 139'-6" 12.3-85 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-16 Radiation Zones (Normal) Compound Building El. 157'-9" & Roof Plan 12.3-86 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-17 Radiation Zones (Normal) Turbine Building El. 73'-0" 12.3-87 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-18 Radiation Zones (Normal ) Turbine Building El. 100'-0" 12.3-88 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-19 Radiation Zones (Normal) CCW HX Building 12.3-89 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-20 Radiation Zones (1 Hour after Accident)-Auxiliary/Containment Building El. 55'-0" 12.3-90 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-21 Radiation Zones (1 Day after Accident)-Auxiliary/Containment Building El. 55'-0" 12.3-91 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-22 Radiation Zones (1 Week after Accident)-Auxiliary/Containment Building El. 55'-0" 12.3-92 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-23 Radiation Zones (1 Month after Accident)-Auxiliary/Containment Building El. 55'-0" 12.3-93 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-24 Radiation Zones (1 Hour after Accident)-Auxiliary/Containment Building El. 78'-0" 12.3-94 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-25 Radiation Zones (1 Day after Accident)-Auxiliary/Containment Building El. 78'-0" 12.3-95 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-26 Radiation Zones (1 Week after Accident)-Auxiliary/Containment Building El. 78'-0" 12.3-96 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-27 Radiation Zones (1 Month after Accident)-Auxiliary/Containment Building El. 78'-0" 12.3-97 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-28 Radiation Zones (1 Hour after Accident)-Auxiliary/Containment Building El. 100'-0" 12.3-98 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-29 Radiation Zones (1 Day after Accident)-Auxiliary/Containment Building El. 100'-0" 12.3-99 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-30 Radiation Zones (1 Week after Accident)-Auxiliary/Containment Building El. 100'-0" 12.3-100 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-31 Radiation Zones (1 Month after Accident)-Auxiliary/Containment Building El. 100'-0" 12.3-101 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-32 Radiation Zones (1 Hour after Accident)-Auxiliary/Containment Building El. 120'-0" 12.3-102 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-33 Radiation Zones (1 Day after Accident)-Auxiliary/Containment Building El. 120'-0" 12.3-103 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-34 Radiation Zones (1 Week after Accident)-Auxiliary/Containment Building El. 120'-0" 12.3-104 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-35 Radiation Zones (1 Month after Accident)-Auxiliary/Containment Building El. 120'-0" 12.3-105 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-36 Radiation Zones (1 Hour after Accident)-Auxiliary/Containment Building El. 137'-6" 12.3-106 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-37 Radiation Zones (1 Day after Accident)-Auxiliary/Containment Building El. 137'-6" 12.3-107 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-38 Radiation Zones (1 Week after Accident)-Auxiliary/Containment Building El. 137'-6" 12.3-108 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-39 Radiation Zones (1 Month after Accident)-Auxiliary/Containment Building El. 137'-6" 12.3-109 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-40 Radiation Zones (1 Hour after Accident)-Auxiliary/Containment Building El. 156'-0" 12.3-110 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-41 Radiation Zones (1 Day after Accident)-Auxiliary/Containment Building El. 156'-0" 12.3-111 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-42 Radiation Zones (1 Week after Accident)-Auxiliary/Containment Building El. 156'-0" 12.3-112 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-43 Radiation Zones (1 Month after Accident)-Auxiliary/Containment Building El. 156'-0" 12.3-113 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-44 Radiation Zones (1 Hour after Accident)-Auxiliary/Containment Building El. 174'-0" 12.3-114 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-45 Radiation Zones (1 Day after Accident)-Auxiliary/Containment Building El. 174'-0" 12.3-115 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-46 Radiation Zones (1 Week after Accident)-Auxiliary/Containment Building El. 174'-0" 12.3-116 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-47 Radiation Zones (1 Month after Accident)-Auxiliary/Containment Building El. 174'-0" 12.3-117 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-48 Radiation Zones (1 Hour after Accident)-Auxiliary/Containment Building El. 195'-0" & Roof Plan 12.3-118 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-49 Radiation Zones (1 Day after Accident)-Auxiliary/Containment Building El. 195'-0" & Roof Plan 12.3-119 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-50 Radiation Zones (1 Week after Accident)-Auxiliary/Containment Building El. 195'-0" & Roof Plan 12.3-120 Rev. 2

APR1400 DCD TIER 2 Security-Related Information - Withhold Under 10 CFR 2.390 Figure 12.3-51 Radiation Zones (1 Month after Accident)-Auxiliary/Containment Building El. 195'-0" & Roof Plan 12.3-121 Rev. 2

APR1400 DCD TIER 2 Figure 12.3-52 Sectional View of Fuel Transfer Tube 12.3-122 Rev. 2

APR1400 DCD TIER 2 12.4 Dose Assessment and Minimization of Contamination 12.4.1 Dose Assessment 12.4.1.1 Estimated Annual Occupational Exposures 12.4.1.1.1 ALARA Design Features for Occupational Dose Reduction A review of the operating data indicates that a significant portion of the occupational exposures (from 50 to 75 percent) are related to the exposure to activated corrosion products. Minimization of primary system corrosion and resultant dose rates is the most effective approach to reduce total personnel occupational exposures.

Leakage of fuel rod cladding accounts for the remaining 25 to 50 percent of pressurized water reactor (PWR) occupational exposure. The fuel rod performance of the APR1400 design is expected to reduce fuel leakage to less than 0.1 percent. This feature also is expected to reduce effluent releases and radwaste activity.

It is expected that proper material selection, chemistry control, and improved fuel cladding leakage reduce the annual occupational exposure.

The design features described in Subsection 12.3.1.5 for SGs greatly reduce the time spent in performing maintenance and inspection. This task represents approximately 25 percent of the total dose for the average PWRs.

The dose received during an outage for SG maintenance is dependent on the number of tubes requiring inspection and repair. The APR1400 SG design provides thermal treatment of the SG tubes and the manufacturing technique found to have avoided the known SG tube cracking. SG tubes are fabricated of Alloy 690 instead of Alloy 600.

Alloy 690 has been proven to be less susceptible to SG tube cracking.

The Palo Verde SG tubes are fabricated of Alloy 600 and processed with thermal treatment.

The Palo Verde plants are known to have no tube cracking problems. This has resulted in a significant reduction in the annual dose received during SG maintenance. Palo Verdes average occupational exposure due to SG maintenance during a refueling outage has been 0.40 person-Sv, in contrast to the Duke Power Companys average of 0.62 person-Sv.

Therefore, a factor of 1.6 reduction in annual occupational exposure is expected for SG maintenance based on improved manufacturing techniques and material selection.

12.4-1 Rev. 2

APR1400 DCD TIER 2 Advanced PWRs like the APR1400 have several additional features that are not provided at Palo Verde or the Duke Power Companys units as described in Subsection 12.3.1.5.

These features reduce the time spent in maintaining the SG and are therefore expected to reduce the annual occupational exposure due to steam generator maintenance by a factor of 1.5. In addition, a factor of 2.5 reduction is expected from material selection, chemistry control, and improved fuel performance as described above. Therefore, all of the above advanced PWR design features result in an overall reduction of the annual occupational exposure due to steam generator maintenance by a factor of 6.

The occupational exposure due to RCP maintenance and inspection represents approximately 4 percent of total station exposures. The APR1400 design minimizes the need for seal replacement and reduces the time for performing this task.

Many other features to reduce occupational exposures are applied to the APR1400 design.

Some examples are:

a. Extended fuel cycle

b. Identification of RCS leakage

c. Integrated head assembly

d. Single stud tensioner

e. Permanent refueling pool seal

f. RCS equipment and piping materials These materials in direct contact with the RCS have a low cobalt content.

g. Steam generator tube material This material has an average cobalt content of less than 0.015 wt%.

h. Equipment reliability, maintainability, and accessibility The APR1400 is designed so that adequate spacing is provided around components. This enhances accessibility, maintainability, and inspectability. In addition, reliable equipment is used so that the frequency of maintenance is minimized for dose reduction.

12.4-2 Rev. 2

APR1400 DCD TIER 2

i. Components such as tanks, piping, and instruments are designed to minimize particulate deposition in accordance with NRC RG 8.8 (Reference 1) recommendations. For example, tanks with sloped bottoms are used in the APR1400 design.

j. System flushing and decontamination capability The APR1400 design provides flushing and decontamination capability to lines where particulate deposition is possible, such as resin transfer lines. This enables operators to flush the lines to prevent buildup of radioactive particulates in the lines, which could result in high personnel exposure.

k. Radwaste handling operations Radwaste handling operations are described in detail in Sections 11.2, 11.3, and 11.4. These systems are designed to operate remotely to the extent practicable, and waste streams are segregated to facilitate processing of waste effectively and efficiently and minimize radiation exposure.

l. Isolation of contaminated components and proper shielding The APR1400 design segregates the radioactive systems from the nonradioactive systems. In addition, components in the radioactive systems are separately located in cubicles based on the source strength, the nature of operation, and the frequency of access. The design also provides adequate shielding between components, valves, and pump galleries to enable personnel to perform operational activities and maintenance in areas of lower radiation.

m. Design for recovery from abnormal occurrences

1) Divisional separation The APR1400 design provides separation of SSCs, which meets the single-failure criterion. Fire and flood barriers are provided to prevent the spread of such events and the loss of equipment in adjacent areas. This design approach also facilitates speedier recovery from fire or flood events.

2) RCS depressurization 12.4-3 Rev. 2

APR1400 DCD TIER 2 Use of the IRWST enables the operators to depressurize the RCS by discharging the coolant to the IRWST through the POSRVs. This tank allows the dissipation of a significant portion of energy to the IRWST, thus facilitating the recovery from loss of heat sink events.

3) Safety injection The provision of a direct header from safety injection, which takes suction from the IRWST to the RV, reduces the likelihood of a failure to provide safety injection water to the reactor.

4) Safety-related equipment power supply The APR1400 design provides Class 1E and redundant electric power to safety-related equipment, thus reducing the probability of safety-related equipment failure due to a loss of power.

These features are further expected to reduce occupational radiation exposures to ALARA levels.

12.4.1.1.2 Occupational Dose Assessment The expected annual cumulative occupational radiation exposure (ORE) from the APR1400 operation is estimated according to the guidance in NRC RG 8.19 (Reference 2). The APR1400 ORE estimation is based on operational experience data from the OPR1000 plant, which is the Korean PWR reactor rated at 1,000 MWe. The design features of this Combustion Engineering type OPR1000 are similar to those of the APR1400.

The dose rates for the APR1400 are conservatively assumed to increase in proportion to the thermal power increase from the OPR1000. During a refueling outage, the ORE is strongly dependent on the activated corrosion products. Although the deposition of the activated corrosion products in the APR1400 is expected to be less because of design improvements, the reduction is not taken into account in the ORE estimation.

According to NRC RG 8.19, the ORE assessment should be based on anticipated radiation conditions after at least 5 years of plant operation. The measured ORE data for a rolling 10-year period from 2004 to 2013 for the OPR1000 (Hanul Unit 3), which started commercial operation in August 1998, are used in the ORE assessment. The internal and 12.4-4 Rev. 2

APR1400 DCD TIER 2 external doses due to airborne radionuclides are negligible because the measurement data for the OPR1000 and NUREG-0713 (Reference 3) demonstrate that the ORE contribution from airborne radioactivity is minimal.

The estimated annual occupational radiation exposures for the APR1400 are determined within the following work categories:

a. Reactor operations and surveillance

b. Routine maintenance

c. Inservice inspection

d. Special maintenance

e. Waste processing

f. Refueling The estimated result of occupational exposure for the APR1400 is presented in Table 12.4-7.

12.4.1.1.2.1 Reactor Operations and Surveillance The performance of various systems and components is monitored by operators. The inspection includes routine inspections and performance tests such as system leakage monitoring, control of instrumentation, electrical and mechanical equipments, and preventive maintenance. Health physics surveys are also included. In some cases, operation of some manual valves requires personnel to enter radiologically controlled areas (RCAs) for short periods.

Examples of routine operation and surveillance activities are as follows:

a. Routine checking and patrol

b. System operation and checking

c. Instrumentation inspections and calibration

d. Unidentified leak checking

e. Communication system checking 12.4-5 Rev. 2

APR1400 DCD TIER 2 These activities may be conducted in the reactor containment, auxiliary building, compound building, and turbine building. The frequency and duration of activities, which are the major factors for ORE dose, are dependent on the operational programs and procedures.

Improvement of the plant layout, access provisions, and operational procedures can reduce exposure time in RCAs for surveillance, inspection, and testing works, thereby minimizing occupational radiation exposure. Details of the design features to minimize ORE are described in Subsection 12.3.1.

Table 12.4-1 provides a breakdown of the collective doses for reactor operations and surveillance activities.

12.4.1.1.2.2 Routine Maintenance Routine maintenance is required for mechanical and electrical components throughout the plant. Maintenance includes routine scheduled maintenance and inspection activities.

Occupational exposure can be reduced by reducing the number of components/equipment that require maintenance and providing ease of maintenance, accessibility of equipment, and ample workspace.

Examples of routine maintenance activities are as follows:

a. Health physics activities

b. Washing and decontamination

c. Leak testing (LLRT, ILRT)

d. Filter & resin replacement

e. Inspection of components

- RCS pumps

- Primary circuit

- Other pumps

- Other valves 12.4-6 Rev. 2

APR1400 DCD TIER 2

- Other mechanical components

- Electrical components

f. Ventilation and filtration system

g. Insulation/shielding and support

h. Miscellaneous works Table 12.4-2 provides a breakdown of the collective doses for routine maintenance activities.

12.4.1.1.2.3 Inservice Inspection ASME Code Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components (Reference 4) requires extensive periodic inservice inspection of various safety-related systems/components and associated welds. These inservice inspections are generally performed during refueling outage.

Examples of inservice inspection activities are as follows:

a. Pressure test

b. Visual inspections

c. Nondestructive and ultrasonic examinations

d. Welds examinations

e. Installation and removal of thermal insulation, platform, etc., and cleanup Table 12.4-3 provides a breakdown of the collective doses for inservice inspection activities.

12.4.1.1.2.4 Special Maintenance Maintenance beyond routine scheduled maintenance is considered special maintenance and can be performed only during refueling outages. This category includes the modification of equipment to upgrade the plant and repairs to any failed components. For special maintenance activities that lead to significant occupational exposure, ORE reduction can be achieved through improved reliability of components and equipments, ease of maintenance 12.4-7 Rev. 2

APR1400 DCD TIER 2 or replacement, use of temporary shielding, improvement of maintenance procedures, and use of remotely operated equipment and robotics.

Examples of special maintenance activities are as follows:

a. Steam generator maintenance

b. Repairs on valves, pumps, heat exchangers, tanks, and other components

c. Pressurizer maintenance Table 12.4-4 provides a breakdown of the collective doses for special maintenance activities.

12.4.1.1.2.5 Waste Processing Radioactive wastes, except spent fuels, are processed by the LWMS, SWMS, and GWMS in the compound building. Radioactive waste processing activities include processing, collecting, storage, and handling of radioactive liquids, solid, and gaseous wastes generated during plant operation. The systems are operated remotely to minimize the expected occupational exposure.

Table 12.4-5 provides a breakdown of the collective doses for waste processing activities.

12.4.1.1.2.6 Refueling Refueling replaces depleted spent fuels with new fuels on an 18-month fuel cycle. This includes receipt and inspection of new fuel assemblies, transfer of fuel assemblies, and replacement of control rods or in-core components. Reactor vessel (RV) opening/closure activities are also included because the RV head must be disassembled and removed before commencing refueling operations. Most of the refueling activities, including cask loading operations, are conducted underwater to provide shielding and maintain occupation dose ALARA. Occupational exposure is also reduced by using automatic cleaning equipment, remote fuel handling, and an integrated head assembly (IHA).

Examples of refueling activities are as follows:

a. RV opening/closure

b. Refueling 12.4-8 Rev. 2

APR1400 DCD TIER 2

c. Stud bolt hole cleaning Table 12.4-6 provides a breakdown of the collective doses for refueling activities.

12.4.1.2 Post-Accident Actions Following a DBA, the vital functions can be performed with an operator radiation exposure of no more than 50 mSv total effective dose equivalent (TEDE) in accordance with the alternative source term (AST) approach, 10 CFR 50.34(f)(2)( vii), GDC 19, NUREG-0737 (Reference 5), Item II.B.2, and plant operating and emergency procedures. For the radiation exposure assessment presented in this section and summarized in Table 12.4-8, the radiation exposures resulting from operator actions required during post-LOCA conditions are evaluated as the bounding DBA for post-accident radiation exposures.

The areas requiring continuous occupancy or infrequent access to perform post-accident vital functions are described in Subsection 12.3.1.8.

All vital areas requiring infrequent access are located in various rooms at different floor elevations in the auxiliary building (AB). The only significant submergence radiation dose that an individual would receive in accessing (transit time) and occupying (stay time) areas that are infrequently accessed is that from the airborne source. The plant emergency procedure mandates the use of respiratory protection to limit intake of airborne radioactive material when the potential for a radiation hazard of 1 DAC exists for the airborne activity.

The assigned protection factor (APF) of 10,000 for a positive pressure self-contained breathing apparatus (SCBA) is used to calculate the inhaled dose rates for the areas requiring infrequent access. 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 12.4(4)). The respirator protections are not provided for areas requiring continuous occupancy, such as the main control room (MCR) and technical support center (TSC).

12.4-9 Rev. 2

APR1400 DCD TIER 2 The dose contribution from the post-LOCA containment low-volume purge system release is negligible. Therefore, the whole body (WB), which is the deep dose equivalent (DDE),

dose rates and the TEDE dose rates due to the post-LOCA containment and ESF leakages are calculated for radiation exposures beginning at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (1 day) using AST methodology.

The radiation exposures from the containment shine are negligible because most of the vital areas requiring infrequent access are heavily self-shielded by concrete structures of the auxiliary building and containment building. However, the Class 1E switchgear rooms and I&C equipment rooms which are located adjacent to the containment building have a minimum concrete shielding of just the containment cylindrical wall of 1.2 m (4 ft).

Therefore, the post-LOCA containment shine dose rates are only considered in these areas.

The MCR and TSC are identified as the areas requiring continuous occupancies to perform the post-accident vital functions. These areas are within the control room (CR) envelope (CRE) boundary. The post-LOCA doses in the CRE are listed in Table 6.4-2. To calculate the CR average dose rate, the CR operator occupancy time based on a 30-day accident duration and CR occupancy factors contained in NRC RG 1.183 (Reference 6) are considered.

The mission doses and dose rates for areas requiring continuous occupancy are listed in Table 12.4-8. The results show that the doses and dose rates meet the criteria of 50 mSv TEDE for 30 days and 0.15 mSv/hr averaged over 30 days as specified in NUREG-0737, Item II.B.2.

The mission dose and dose rates for areas requiring infrequent access to perform post-accident vital functions are also presented in Table 12.4-8. The estimate of the access times is based on a walk speed of 4 km/h (13000 ft/h) for horizontal pathways and 2 km/h (6500 ft/h) where upstairs/downstairs are included. It is assumed that 20% margin is added to the transit time and the direct dose to the vital area mission pathways is doubled to account for any potential nonconservatism due to simplifications in the calculations.

Table 12.4-8 indicates that the doses to the plant personnel calculated based on anticipated stay and access times at various areas at various times are within the criterion of 50 mSv TEDE as specified in NUREG-0737, Item II.B.2.

12.4.1.3 Estimated Annual Dose at the Exclusion Area Boundary 12.4-10 Rev. 2

APR1400 DCD TIER 2 The direct radiation from the reactor containment, auxiliary, compound, and turbine generator buildings is negligible compared with that from outside storage tanks. These tanks are expected to contain radioactivity that yields a dose rate of 0.0025 mSv/hr or less at the tank shielded surface. Therefore, the estimated annual direct dose at the site boundary is expected to be small.

The estimated doses at the site boundary due to released activity are given in Subsections 11.2.3.1 and 11.3.3.1. The resultant annual doses for the general public are within the limits of the applicable regulations.

12.4.1.4 Estimated Dose to Construction Workers The construction worker doses are the sum of the exposures due to direct radiation and airborne radiation from the adjacent operating unit(s). The dose criterion to the construction workers is 1 mSv/yr. The construction worker doses are site-specific and based on the number of operating units, distances, meteorological conditions, and construction schedule. 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(1)).

12.4.2 Minimization of Contamination and Radioactive Waste Generation The APR1400 design includes features and operational programs to conform with 10 CFR 20.1406 and NRC RG 4.21 (Reference 7) to minimize contamination of the facility and environment, facilitate decommissioning, and minimize the generation of radioactive waste throughout the life cycle of the facility.

The design includes control measures that address the requirements in Regulatory Positions C.1 through C.4 and Appendix A of NRC RG 4. 21. The control measures are designed to work in conjunction with the operational procedures for the collection, processing, sampling, storing, monitoring, and disposal of radioactive waste during normal operation, including AOOs. 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 (2)). The design also takes into account the life-cycle planning of the facility and through COL 12.4(2), identifies the operational programs and procedures to address maintenance of structures, systems, and components (SSCs), site radiological environmental monitoring, 12.4-11 Rev. 2

APR1400 DCD TIER 2 and documentation of operational incidents for decommissioning planning associated with NRC RG 4.21. The APR1400 systems with the potential to contain radiological contamination are listed in Table 12.4-9 and require full NRC RG 4.21 evaluations.

Because NRC RG 4.21 control measures cover different phases of the plant life cycle, the APR1400 design divides the control measures into the following six design and operational objectives that are used to evaluate the conformance of the measures with 10 CFR 20.1406 and NRC RG 4.21:

a. Objective 1 - Prevention/Minimization of Unintended Contamination

b. Objective 2 - Provision of Adequate and Early Leak Detection Capability

c. Objective 3 - Reduction of Cross-Contamination, Decontamination, and Waste Generation

d. Objective 4 - Decommissioning Planning

e. Objective 5 - Operations and Documentation

f. Objective 6 - Site Radiological Environmental Monitoring The design features in the systems that are intended to meet the objectives for minimizing contamination and the generation of radioactive waste are described in Table 12.4-10. The table also provides cross-references to the DCD sections that provide further information on how the SSCs meet the requirements of 10 CFR 20.1406 and NRC RG 4.21.

12.4.2.1 Design Considerations The control measures delineated in Regulatory Positions C.1 through C.4 and Appendix A of NRC RG 4.21 are categorized into four design objectives that address the design features described below.

Objective 1 - Prevention/Minimization of Unintended Contamination Objective 1 requires control measures to prevent and/or minimize radioactive contamination of the facility and environment. This objective is met by implementing design features such as:

12.4-12 Rev. 2

APR1400 DCD TIER 2

a. Selection of proper materials for SSCs to withstand process conditions without the generation of excess contamination, such as corrosion-resistant piping

b. Use of containment walls or dikes around contaminated or potentially contaminated SSCs and other means of prevention/minimization such as leak detection, drain collection, double-walled piping, and piping sleeves with leak detection for piping between buildings

c. Minimization of buried or embedded piping and drains (alternatively utilizing piping within a concrete structure or double-walled piping, both with leak detection capabilities)

d. Provisions for proper operational interlocks and alarms and administrative controls to avoid inadvertent bypasses and releases

e. Minimization of the potential for component failures through the use of nuclear-industry-proven technologies, reinforced quality control, and strict adherence to the applicable codes and standards Objective 2 - Provision of Adequate and Early Leak Detection Capability Objective 2 requires provisions for leak detection capabilities, allowing prompt assessment to support a timely and appropriate response in the event of an unintended release of radioactive contamination. Leak detection systems are included within the facility design to initiate alarm signals in the radwaste control room (where applicable) and MCR for timely operator actions. This objective is met by the implementing the following design features:

a. Adequate leak detection instruments (including individual and separate leak detection instruments for plant SSCs with a potential for leakage to the extent possible)

b. Facility design (e.g., floors that slope toward drainage collection piping, level detection instruments, smooth surfaces for drainage, catch basins)

c. Drainage collection and pathways and the placement of detection instruments as close as practicable to the sources of leaks 12.4-13 Rev. 2

APR1400 DCD TIER 2

d. Sensitive (including smaller liquid volume collection) liquid/moisture detection instruments

e. Facilitation of periodic calibration and replacement of leak detection instruments

f. Adequate access pathways and spaces for prompt assessment of detected leakage and appropriate operator responses, tank overflow management, and adequate sampling points for operational analysis Objective 3 - Reduction of Cross-Contamination, Decontamination, and Waste Generation Objective 3 is met by providing capabilities to reduce cross-contamination, the need for decontamination, and the generation of radioactive waste. The control measures that are implemented to meet this objective include:

a. Segregation of components in accordance with contamination types, levels, and characteristics

b. Onsite decontamination facility

c. Flushing capabilities for components and piping that handle or transport radioactive fluid

d. Smooth and cleanable surfaces for component fabrication and facility structures such as epoxy coating for cubicle floors

e. Capability to temporarily hold and isolate contamination and quickly initiate operator action to provide timely and appropriate responses

f. Life-cycle planning of waste processing, packaging, handling, storing, and shipment for disposal Objective 4 - Decommissioning Planning Objective 4 is met by providing features that facilitate decommissioning. The design features include:

a. Modular construction 12.4-14 Rev. 2

APR1400 DCD TIER 2

b. Minimization of buried and embedded components and piping

c. Documentation and information in a centralized area for ready recovery

d. Removable block walls to facilitate the removal of SSCs

e. SSCs that minimize the generation of radioactive waste during decommissioning activities 12.4.2.2 Operational/Programmatic Considerations The operational programs and procedures in 10 CFR 20.1406 and 10 CFR 20.1501 must be met. The COL applicant is to address Objectives 5 and 6, described below, to meet the operational and programmatic requirements of NRC RG 4.21 Positions C.1 through C.4 and RG 4.22 (Reference 8), as applicable (COL 12.4(2)).

The requirements are related to procedures and programs that support the operation and maintenance of the design features described in Objectives 1 through 4 and documentation of any incidents that may occur during the operating life of the facility. The documentation of these incidents, including spillage, leakage, overflows, and associated cleanup requirements, is an integral part of decommissioning planning.

Objective 5 - Operations and Documentation Objective 5 is met by developing operational procedures for the following:

a. Supplemental control of the SSCs to prevent/minimize the spread of contamination

b. Periodic review of operational procedures to provide reasonable assurance that the procedures reflect lessons learned and up-to-date content, including personnel qualification and training

c. Accurate records and documentation of design, construction, modifications, and operational incidences to facilitate decommissioning

d. Development of radwaste processing and management programs 12.4-15 Rev. 2

APR1400 DCD TIER 2

e. Development of maintenance and surveillance programs in accordance with NRC RG 4.21 and RG 4.22, as applicable, to provide reasonable assurance that SSCs will perform as designed to prevent the spread of contamination

f. Documentation of the results of subsurface radiological surveys in accordance with 10 CFR 20.1501 Objective 6 - Site Radiological Environmental Monitoring Objective 6 is met by developing a conceptual site model (COL 12.4 (2)) that:

a. Includes site-specific hydrogeology, potential migration, and groundwater transport pathways

b. Assesses the effects of construction on hydrogeological characteristics of the site

c. Establishes a site-specific contamination monitoring program for potential groundwater pathways from release source to receptor point, in accordance with 10 CFR 20.1406 and the guidance of RG 4.21 12.4.2.3 Summary of Design Features The APR1400 design features that minimize facility and environmental contamination are described in the relevant sections of this DCD. Table 12.4-10 summarizes how these features meet the design objectives in accordance with the requirements of 10 CFR 20.1206 and NRC RG 4.21 and provides cross-references to the relevant DCD subsections.

Many of the features are designed to work together to support the defense-in-depth strategy for the prevention and minimization of the contamination of the facility and the environment.

12.4.2.4 Design Details to Address Minimization of Contamination 12.4.2.4.1 Early Leak Detection Drain Pipe The early leak detection design for plant areas containing radioactive components consists of a drain pipe with a level switch and a manual valve installed on the exposed pipe portion in the valve pit or corresponding location of the trench. The drain pipe collects drainage from cubicles that may contain radioactive liquid from leakage or overflow. The drain 12.4-16 Rev. 2

APR1400 DCD TIER 2 pipe is a 5.08 cm (2 in.) stainless steel pipe embedded in the basemat and is sloped to facilitate drainage. The embedded portion of the drain pipe is encased in an outer pipe.

The inner pipe is open to drainage into the trench, which directs the collected liquid to the normal sump for collection. Because the drain pipe is relatively small, a small leak can be detected, thus achieving early detection. The level switch is designed to initiate an alarm in the MCR for operator action. Early leak detection capability is built into each cubicle that may contain radioactive or potentially radioactive fluid for quick assessment. The typical design of the leak detection configuration is presented in Figure 12.4-1. The leak detection instrumentation is shown in Figure 12.4-2.

12.4.2.4.2 Leak Collection Trench Concrete trenches are provided within the basemat of the auxiliary and compound buildings to facilitate leak drainage to liquid collection sumps. The trench has an epoxy coating and is significantly sloped to facilitate flow and prevent the accumulation of fluid along the flow path. The trench has removable steel grated covers to facilitate access for inspection and calibration of leak detection instruments; trench cleaning; and inspection, repair, and maintenance of the epoxy coating. Where area drainage is expected to contain significant debris, the trench is equipped with screens or a drain pipe to facilitate liquid flow without entraining solids or debris in the sumps. The typical design of the compound building trench design is presented in Figure 12.4-3.

12.4.2.4.3 Penetration Design between Buildings Piping that extends from one building to another with a separation gap between the buildings is equipped with sleeves or double-walled concentric piping to prevent direct, unintended leakage to the environment. The use of either piping sleeves or double-walled concentric piping depends on the separation distances, with piping sleeves generally used for smaller separation gaps and double-walled piping used for larger separation gaps.

Piping sleeves and the outer wall piping for the double-walled piping are equipped with seals to maintain building pressure differentials and to prevent the infiltration of outside water and air into the buildings. Piping sleeves and outer piping are also equipped with sleeve joints to accommodate some degree of building movement. Piping sleeves and outer wall piping are installed with a mild slope to facilitate drainage of liquid in the event of inner pipe leakage. The leakage is drained directly into a nearby sump. The typical 12.4-17 Rev. 2

APR1400 DCD TIER 2 designs for piping penetrations between buildings and building sleeves are presented in Figure 12.4-4.

12.4.2.4.4 Minimization of Embedded and/or Buried Piping Embedded or buried piping is minimized using the following approach:

a. Piping is routed inside pipe chases to the extent possible to minimize embedded piping. Drain pipes on the floors above the basemat level are allowed to penetrate the floor and are routed inside pipe chases below the floor level as required.

b. To the extent practicable, double-walled piping is used when embedded piping segments cannot be avoided. The outside piping of the double wall pipe is designed to drain to a local sump.

c. For the basemat level, drain pipes are routed in concrete trenches that are sloped toward the local sump. The trenches are coated with epoxy to facilitate drainage and cleaning.

d. For piping that carries contaminated or potentially contaminated fluid and is located outside the plant structures, the piping is routed in underground concrete tunnels to the maximum extent practicable. Systems that contain yard piping routed through underground concrete tunnels may include CCWS, ESWS, CVCS, condensate systems, etc., which are determined based upon site-specific plant layout conditions. The COL applicant is to implement concrete tunnels for those systems that 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 (COL 12.4(3)).

12.4.3 Combined License Information COL 12.4(1) The COL applicant is to estimate construction worker doses based on the site-specific information, such as the number of operating units, distances from radiation sources, meteorological conditions, and construction schedule.

12.4-18 Rev. 2

APR1400 DCD TIER 2 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.

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.

12.4.4 References

1. Regulatory Guide 8.8, Information Relevant to Ensuring the Occupational Radiation Exposures at Nuclear Power Stations will be ALARA, Rev. 3, U.S. Nuclear Regulatory Commission, June 1978.

2. Regulatory Guide 8.19, Occupational Radiation Dose Assessment in Light Water Reactor Power Plants Design Stage Man-Rem Estimates, Rev. 1, U.S. Nuclear Regulatory Commission, June 1979.

3. NUREG-0713, Occupational Radiation Exposure at Commercial Nuclear Power Reactors and Other Facilities, U.S. Nuclear Regulatory Commission.

12.4-19 Rev. 2

APR1400 DCD TIER 2

4. ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, The American Society of Mechanical Engineers, 2007.

5. NUREG-0737, Clarification of TMI Action Plan Requirements, U.S. Nuclear Regulatory Commission, November 1980.

6. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, U.S. Nuclear Regulatory Commission, July 2000.

7. Regulatory Guide 4.21, Minimization of Contamination and Radioactive Waste Generation: Life-Cycle Planning, U.S. Nuclear Regulatory Commission, June 2008.

8. Regulatory Guide 4.22, Decommissioning Planning During Operations, U.S.

Nuclear Regulatory Commission, December 2012.

12.4-20 Rev. 2

APR1400 DCD TIER 2 Table 12.4-1 Occupational Dose Estimates During Reactor Operations and Surveillance Average Exposure Annual Dose Rate Number of Time (1) ORE Dose Activity [Sv/hr] Workers (1) Frequency (1) [hr] [mSv/yr]

Routine checking and 1.5 2 Per Shift 1.3 4.3 patrol System operation and 1.1 3 Per Shift 3.4 12.3 checking Instrumentation 6.4 3 Daily 2.7 18.9 inspections and calibration Unidentified leak 96.1 1 Monthly 0.8 0.9 checking Communication system 1.0 1 Daily 0.6 0.2 checking Total 36.6 (1) In order to present statistical information on routine and non-routine activities, values for these parameters (number of workers, frequency and exposure time) are adjusted so that the multiplication of these parameters by the average dose rate equals the annual ORE dose, which is the measured value at Hanul Unit 3.

12.4-21 Rev. 2

APR1400 DCD TIER 2 Table 12.4-2 Occupational Dose Estimates During Routine Maintenance Average Exposure Annual Dose Rate Number of Time (1) ORE Dose Activity [Sv/hr] Workers (1) Frequency (1), (2) [hr] [mSv/yr]

Health physics activities 1.6 8 Daily 6.4 29.9 Washing and decontamination 3.9 4 Daily 1.5 8.5 Leak testing (LLRT, ILRT) 3.6 4 Monthly 11.8 2.0 Filter & resin replacement 46.7 4 Monthly 3.6 8.1 RCS pumps 12.9 10 Per Cycle 299.0 27.0 Primary circuit 19.2 10 Per Cycle 220.2 29.6 Other pumps 2.9 5 Per Cycle 119.8 1.2 Inspection of Other valves 14.1 5 Per Cycle 257.6 12.7 components Other mechanical 1.0 5 Per Cycle 207.3 0.7 components Electrical 0.6 3 Per Cycle 51.8 0.1 components Ventilation and filtration system 3.9 8 Monthly 6.3 2.4 Insulation/shielding and support 8.2 3 Monthly 3.0 0.9 Miscellaneous works 5.5 7 Daily 1.6 22.5 Total 145.6 (1) In order to present statistical information on routine and non-routine activities, values for these parameters (number of workers, frequency and exposure time) are adjusted so that the multiplication of these parameters by the average dose rate equals the annual ORE dose, which is the measured value at Hanul Unit 3.

(2) For the cases of Per Cycle, an approximate frequency of 0.7 per year is used to calculate the annual ORE dose, since a fuel cycle of 18 months is applied.

12.4-22 Rev. 2

APR1400 DCD TIER 2 Table 12.4-3 Occupational Dose Estimates During Inservice Inspection Average Exposure Annual Dose Rate Number of Time (1) ORE Dose Activity [Sv/hr] Workers (1) Frequency (1), (2) [hr] [mSv/yr]

Inspections of components 26.1 7 Per Cycle 171.3 21.9 and systems Inspection of welds 39.3 3 Per Cycle 15.4 1.3 ISI support (3) 29.0 5 Per Cycle 112.4 11.4 Total 34.6 (1) In order to present statistical information on routine and non-routine activities, values for these parameters (number of workers, frequency and exposure time) are adjusted so that the multiplication of these parameters by the average dose rate equals the annual ORE dose, which is the measured value at Hanul Unit 3.

(2) For the cases of Per Cycle, an approximate frequency of 0.7 per year is used to calculate the annual ORE dose, since a fuel cycle of 18 months is applied.

(3) Activities such as installation and removal of platforms, ladders, thermal insulation, etc. and preparation of work, etc.

12.4-23 Rev. 2

APR1400 DCD TIER 2 Table 12.4-4 Occupational Dose Estimates During Special Maintenance Average Exposure Annual Dose Rate Number of Frequency Time (1) ORE Dose Activity [Sv/hr] Workers (1) (1), (2)

[hr] [mSv/yr]

RCS pump 20.8 5 Per Cycle 161.2 11.7 Primary circuit 47.8 5 Per Cycle 228.1 38.2 Other pumps 5.1 5 Per Cycle 4.0 0.1 Maintenance of Other valves 17.0 5 Per Cycle 42.0 2.5 components Other mechanical components 0.8 5 Per Cycle 52.3 0.1 Electric components 0.5 4 Per Cycle 5.6 0.0 Piping 53.1 4 Per Cycle 12.8 1.9 Maintenance of pressurizer 42.1 10 Per Cycle 37.2 11.0 Man-way opening/closure 70.2 5 Per Cycle 18.3 4.5 Maintenance of Nozzle dam activities 231.5 5 Per Cycle 16.7 13.5 SG primary Tube inspection 51.2 5 Per Cycle 154.2 27.6 side Tube repairs 33.1 5 Per Cycle 243.4 28.2 Other works 31.3 10 Per Cycle 138.9 30.4 Maintenance Man-way opening/closure 17.4 5 Per Cycle 23.3 1.4 of Inspection & cleaning 25.2 8 Per Cycle 68.7 9.7 SG secondary Other works Per Cycle 1.5 10 819.1 8.6 side Servicing of control rod drive mechanism 17.1 3 Per Cycle 265.8 9.5 Total 199.0 (1) In order to present statistical information on routine and non-routine activities, values for these parameters (number of workers, frequency and exposure time) are adjusted so that the multiplication of these parameters by the average dose rate equals the annual ORE dose, which is the measured value at Hanul Unit 3.

(2) For the cases of Per Cycle, an approximate frequency of 0.7 per year is used to calculate the annual ORE dose, since a fuel cycle of 18 months is applied.

12.4-24 Rev. 2

APR1400 DCD TIER 2 Table 12.4-5 Occupational Dose Estimates During Waste Processing Average Exposure Annual Dose Rate Number of Time (1) ORE Dose Activity [Sv/hr] Workers (1) Frequency (1) [hr] [mSv/yr]

System operation, inspection 1.7 1 Weekly 0.7 0.1 and testing Sampling and analysis 0.5 1 Daily 3.0 0.5 Operation of waste processing 4.7 3 Weekly 7.6 5.4 and packaging equipment Total 6.0 (1) In order to present statistical information on routine and non-routine activities, values for these parameters (number of workers, frequency and exposure time) are adjusted so that the multiplication of these parameters by the average dose rate equals the annual ORE dose, which is the measured value at Hanul Unit 3.

12.4-25 Rev. 2

APR1400 DCD TIER 2 Table 12.4-6 Occupational Dose Estimates During Refueling Average Exposure Annual Dose Rate Number of Frequency Time (1) ORE Dose Activity [Sv/hr] Workers (1) (1), (2)

[hr] [mSv/yr]

RPV head removal and 36.9 10 Per Cycle 170.0 43.9 installation Inspection of RPV & 59.0 5 Per Cycle 169.3 35.0 reactor internals Works In-core and Ex-core 55.6 5 Per Cycle 66.6 13.0 related to instrumentation RPV Inspection and works 70.7 5 Per Cycle 71.0 17.6 related threaded holes Reactor clean-up 3.3 5 Per Cycle 2.7 < 0.1 Other works 13.0 15 Per Cycle 209.0 28.5 Refueling preparation 13.9 3 Per Cycle 77.1 2.3 Fuel withdrawal 6.0 5 Per Cycle 125.8 2.6 Works Fuel inspection 1.8 5 Per Cycle 147.7 0.9 related to Fuel loading 6.1 5 Per Cycle 84.6 1.8 fuel handling Refueling machine 34.2 5 Per Cycle 120.1 14.4 inspection &

maintenance Other works 4.3 10 Per Cycle 108.1 3.3 Total 163.2 (1) In order to present statistical information on routine and non-routine activities, values for these parameters (number of workers, frequency and exposure time) are adjusted so that the multiplication of these parameters by the average dose rate equals the annual ORE dose, which is the measured value at Hanul Unit 3.

(2) For the cases of Per Cycle, an approximate frequency of 0.7 per year is used to calculate the annual ORE dose, since a fuel cycle of 18 months is applied.

12.4-26 Rev. 2

APR1400 DCD TIER 2 Table 12.4-7 Annual Personnel Doses per Activity Categories Estimated Annual ORE Dose Fraction Category of Activity [%] [man*mSv]

Reactor operations and surveillance 6.3% 36.6 Routine maintenance 24.9% 145.6 Inservice inspection 5.9% 34.6 Special maintenance 34.0% 199.0 Waste processing 1.0% 6.0 Refueling 27.9% 163.2 Total 585.0 12.4-27 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (1 of 21)

Estimated Accident Mission Dose

a. The vital area mission doses and dose rates for areas requiring continuous occupancy to perform post-accident vital functions Regulatory Limit Area Requiring Continuous Dose Dose Rate Dose Dose Rate Occupancy (mSv) (mSv/hr) (mSv) (mSv/hr)

MCR 46.9 0.148 50 0.15 TSC 46.9 0.148 50 0.15

b. The vital area mission doses and dose rates for access to areas requiring continuous occupancy to perform post-accident vital functions Mission Dose Post-LOCA Dose Rate Occupancy Dose Access Vital Area Action Time (hr) (mSv/hr) Time (min) (mSv) Route 1 1.06 5.44 0.10 Access from 2 5.23 5.44 0.47 Outside Fig. 12.3-MCR/TSC 4 6.73 5.44 0.61 (Comp. 13,28,40 Bldg)(1) 8 4.09 5.44 0.37 24 0.92 5.44 0.08 (1) Section within the row line from PE to PF along the column line 31 in Figure 12.3-13 12.4-28 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (2 of 21)

c. Vital area mission dose for areas requiring infrequent access to perform post-accident vital functions (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after LOCA) (1 of 4)

Mission Dose Vital Area Access Action Dose Rate Occupancy Dose (Room No.) Route (mSv/hr) Time (min) (mSv)

Post-accident Sample Access from Outside 37.51 6.07 3.79 and Control Panel (Comp. Bldg)(1) Fig. 12.3-Room Sampling 1.43 20 0.48 13,20,28 (055-A48B, A49B) Sum 4.27 Access from Post-accident Sample and 35.24 6.47 3.80 Radiochemistry Lab Control Panel Room Fig. 12.3-(085-P37) Analysis 0.91 60 0.91 12,13,20,28 Return to MCR/TSC 1.06 6.45 0.11 Sum 4.83 Access from Post-accident Sample and 33.63 6.79 3.81 Sample Counting Control Panel Room Fig. 12.3-Room Analysis 0.91 60 0.91 12,13,20,28 (085-P36)

Return to MCR/TSC 1.06 6.77 0.12 Sum 4.84 Access from Outside 1.06 5.78 0.10 (Comp. Bldg)(1)

Remote Shutdown Operation 0.91 90 1.37 Fig. 12.3-Room Return to Outside 13,28,36 (137-A06D) 1.06 3.35 0.06 (Aux. Bldg)(2)

Sum 1.53 Access from MCR/TSC 1.24 6.12 0.13 Remote Control Operation 0.91 90 1.37 Fig. 12.3-Console Room Return to MCR/TSC 1.24 6.12 0.13 32,36,40 (137-A41A)

Sum 1.62 (1) Section within the row line from PE to PF along the column line 31 in Figure 12.3-13 (2) Section within the column line from 14 to 15 along the row line AK in Figure 12.3-28 12.4-29 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (3 of 21)

c. Vital area mission dose for areas requiring infrequent access to perform post-accident vital functions (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after LOCA) (2 of 4)

Mission Dose Vital Area Access Action Dose Rate Occupancy Dose (Room No.) Route (mSv/hr) Time (min) (mSv)

Access from 1.91 6.35 0.20 MCR/TSC Class 1E Switchgear Inspection 136.51 10 22.75 Fig. 12.3-01A Room Return to Outside 13,24,28,40 (078-A25A) 10.39 2.78 0.48 (Comp. Bldg)(1)

Sum 23.44 Access from 1.91 6.31 0.20 MCR/TSC Class 1E Switchgear Inspection 139.63 10 23.27 Fig. 12.3-01B Room Return to Outside 24,28,40 (078-A25B) 13.19 2.03 0.45 (Aux. Bldg)(2)

Sum 23.92 Access from 2.04 5.30 0.18 Class 1E Switchgear MCR/TSC Fig. 12.3-01C Room Inspection 0.91 10 0.15 24,40 (078-A02C) Return to MCR/TSC 2.04 5.30 0.18 Sum 0.51 Access from 2.04 5.26 0.18 Class 1E Switchgear MCR/TSC Inspection 0.91 10 0.15 Fig. 12.3-01D Room Return to MCR/TSC 2.04 5.26 0.18 24,40 (078-A02D)

Sum 0.51 Access from Outside 1.06 5.38 0.10 (Comp. Bldg)(3)

I&C Equip. Room Inspection 18.40 30 9.20 Fig. 12.3-(157-A19C) Return to Outside 13,28,40 1.06 4.10 0.07 (Aux. Bldg)(4)

Sum 9.37 (1) Section within the row line from PI to PJ along the column line 31 in Figure 12.3-13 (2) Section within the column line from 23 to 24 along the row line AK in Figure 12.3-28 (3) Section within the row line from PE to PF along the column line 31 in Figure 12.3-13 (4) Section within the column line from 14 to 15 along the row line AA in Figure 12.3-28 12.4-30 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (4 of 21)

c. Vital area mission dose for areas requiring infrequent access to perform post-accident vital functions (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after LOCA) (3 of 4)

Mission Dose Vital Area Access Action Dose Rate Occupancy Dose (Room No.) Route (mSv/hr) Time (min) (mSv)

Access from Outside 1.06 5.31 0.09 (Comp. Bldg)(1)

I&C Equip. Room Inspection 0.91 30 0.46 Fig. 12.3-(157-A25C) Return to Outside 13,28,40 1.06 4.04 0.07 (Aux. Bldg)(2)

Sum 0.62 Access from Outside 1.06 6.29 0.11 (Comp. Bldg)(1)

I&C Equip. Room Inspection 0.91 30 0.46 Fig. 12.3-(157-A01D) Return to Outside 13,28,40 1.06 3.87 0.07 (Aux. Bldg)(3)

Sum 0.64 Access from Outside 1.06 6.52 0.12 (Comp. Bldg)(1)

I&C Equip. Room Inspection 18.40 30 9.20 Fig. 12.3-(157-A19D) Return to Outside 13,28,40 1.06 4.09 0.07 (Aux. Bldg)(3)

Sum 9.39 Access from 17.62 9.41 2.76 MCR/TSC Access Area outside Operation 1.74 30 0.87 Fig. 12.3-the CS Pump Room Return to Outside 13,20,28,40 (050-A01C) 32.71 4.93 2.69 (Comp. Bldg)(4)

Sum 6.32 (1) Section within the row line from PE to PF along the column line 31 in Figure 12.3-13 (2) Section within the column line from 14 to 15 along the row line AA in Figure 12.3-28 (3) Section within the column line from 14 to 15 along the row line AK in Figure 12.3-28 (4) Section within the row line from PI to PJ along the column line 31 in Figure 12.3-13 12.4-31 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (5 of 21)

c. Vital area mission dose for areas requiring infrequent access to perform post-accident vital functions (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after LOCA) (4 of 4)

Mission Dose Vital Area Access Action Dose Rate Occupancy Dose (Room No.) Route (mSv/hr) Time (min) (mSv)

Access from Outside 35.53 6.47 3.83 (Comp. Bldg)(1)

Access Area outside Operation 1.74 30 0.87 Fig. 12.3-the CS Pump Room Return to Outside 13,20,28 (050-A01D) 37.09 4.11 2.54 (Aux. Bldg)(2)

Sum 7.24 Access from 18.72 8.77 2.74 MCR/TSC Access Area outside Operation 1.74 30 0.87 Fig. 12.3-the SC Pump Room Return to Outside 13,20,28,40 (050-A04A) 37.23 4.28 2.66 (Comp. Bldg)(3)

Sum 6.26 Access from Outside 39.17 5.82 3.80 (Comp. Bldg)(1)

Access Area outside Operation 1.74 30 0.87 Fig. 12.3-the SC Pump Room Return to Outside 13,20,28 (050-A04B) 43.98 3.43 2.51 (Aux. Bldg)(2)

Sum 7.18 (1) Section within the row line from PE to PF along the column line 31 in figure 12.3-13 (2) Section within the column line from 23 to 24 along the row line AK in Figure 12.3-28 (3) Section within the row line from PI to PJ along the column line 31 in Figure 12.3-13 12.4-32 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (6 of 21)

d. Vital area mission dose for areas requiring infrequent access to perform post-accident vital functions (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after LOCA) (1 of 4)

Mission Dose Vital Area Access Action Dose Rate Occupancy Dose (Room No.) Route (mSv/hr) Time (min) (mSv)

Post-accident Sample Access from Outside 41.67 6.07 4.22 and Control Panel (Comp. Bldg)(1) Fig. 12.3-Room Sampling 5.60 20 1.87 13,20,28 (055-A48B, A49B) Sum 6.08 Access from Post-accident Sample and 39.40 6.47 4.25 Radiochemistry Lab Control Panel Room Fig. 12.3-(085-P37) Analysis 5.08 60 5.08 12,13,20,28 Return to MCR/TSC 5.23 6.45 0.56 Sum 9.89 Access from Post-accident Sample and 37.79 6.79 4.28 Sample Counting Control Panel Room Fig. 12.3-Room Analysis 5.08 60 5.08 12,13,20,28 (085-P36)

Return to MCR/TSC 5.23 6.77 0.59 Sum 9.94 Access from Outside 5.23 5.78 0.50 (Comp. Bldg)(1)

Remote Shutdown Operation 5.08 90 7.61 Fig. 12.3-Room Return to Outside 13,28,36 (137-A06D) 5.23 3.35 0.29 (Aux. Bldg)(2)

Sum 8.41 Access from MCR/TSC 6.23 6.12 0.64 Remote Control Operation 5.08 90 7.61 Fig. 12.3-Console Room Return to MCR/TSC 6.23 6.12 0.64 32,36,40 (137-A41A)

Sum 8.89 (1) Section within the row line from PE to PF along the column line 31 in Figure 12.3-13 (2) Section within the column line from 14 to 15 along the row line AK in Figure 12.3-28 12.4-33 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (7 of 21)

d. Vital area mission dose for areas requiring infrequent access to perform post-accident vital functions (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after LOCA) (2 of 4)

Mission Dose Vital Area Access Action Dose Rate Occupancy Dose (Room No.) Route (mSv/hr) Time (min) (mSv)

Access from 6.07 6.35 0.64 MCR/TSC Class 1E Switchgear Inspection 140.68 10 23.45 Fig. 12.3-01A Room Return to Outside 13,24,28,40 (078-A25A) 14.55 2.78 0.68 (Comp. Bldg)(1)

Sum 24.76 Access from 6.08 6.31 0.64 Class 1E Switchgear MCR/TSC Fig. 12.3-01B Room Inspection 143.80 10 23.97 24,40 (078-A25B) Return to MCR/TSC 6.08 6.31 0.64 Sum 25.24 Access from 6.20 5.30 0.55 Class 1E Switchgear MCR/TSC Fig. 12.3-01C Room Inspection 5.08 10 0.85 24,40 (078-A02C) Return to MCR/TSC 6.20 5.30 0.55 Sum 1.94 Access from 6.21 5.26 0.54 Class 1E Switchgear MCR/TSC Inspection 5.08 10 0.85 Fig. 12.3-01D Room Return to MCR/TSC 6.21 5.26 0.54 24,40 (078-A02D)

Sum 1.94 Access from Outside 5.23 5.38 0.47 (Comp. Bldg)(2)

I&C Equip. Room Inspection 22.57 30 11.28 Fig. 12.3-(157-A19C) Return to Outside 13,28,40 5.23 4.10 0.36 (Aux. Bldg)(3)

Sum 12.11 (1) Section within the row line from PI to PJ along the column line 31 in Figure 12.3-13 (2) Section within the row line from PE to PF along the column line 31 in Figure 12.3-13 (3) Section within the column line from 14 to 15 along the row line AA in Figure 12.3-28 12.4-34 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (8 of 21)

d. Vital area mission dose for areas requiring infrequent access to perform post-accident vital functions (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after LOCA) (3 of 4)

Mission Dose Vital Area Access Action Dose Rate Occupancy Dose (Room No.) Route (mSv/hr) Time (min) (mSv)

Access from Outside 5.23 5.31 0.46 (Comp. Bldg)(1)

I&C Equip. Room Inspection 5.08 30 2.54 Fig. 12.3-(157-A25C) Return to Outside 13,28,40 5.23 4.04 0.35 (Aux. Bldg)(2)

Sum 3.35 Access from Outside 5.23 6.29 0.55 (Comp. Bldg)(1)

I&C Equip. Room Inspection 5.08 30 2.54 Fig. 12.3-(157-A01D) Return to Outside 13,28,40 5.23 3.87 0.34 (Aux. Bldg)(3)

Sum 3.42 Access from Outside 5.23 6.52 0.57 (Comp. Bldg)(1)

I&C Equip. Room Inspection 22.57 30 11.28 Fig. 12.3-(157-A19D) Return to Outside 13,28,40 5.23 4.09 0.36 (Aux. Bldg)(3)

Sum 12.21 Access from 21.79 9.41 3.42 MCR/TSC Access Area outside Operation 5.90 30 2.95 Fig. 12.3-the CS Pump Room Return to Outside 13,20,28,40 (050-A01C) 36.87 4.93 3.03 (Comp. Bldg)(4)

Sum 9.40 (1) Section within the row line from PE to PF along the column line 31 in Figure 12.3-13 (2) Section within the column line from 14 to 15 along the row line AA in Figure 12.3-28 (3) Section within the column line from 14 to 15 along the row line AK in Figure 12.3-28 (4) Section within the row line from PI to PJ along the column line 31 in Figure 12.3-13 12.4-35 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (9 of 21)

d. Vital area mission dose for areas requiring infrequent access to perform post-accident vital functions (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after LOCA) (4 of 4)

Mission Dose Vital Area Access Action Dose Rate Occupancy Dose (Room No.) Route (mSv/hr) Time (min) (mSv)

Access from Outside 39.69 6.47 4.28 (Comp. Bldg)(1)

Access Area outside Operation 5.90 30 2.95 Fig. 12.3-the CS Pump Room Return to Outside 13,20,28 (050-A01D) 41.25 4.11 2.83 (Aux. Bldg)(2)

Sum 10.06 Access from 22.89 8.77 3.34 MCR/TSC Access Area outside Operation 5.90 30 2.95 Fig. 12.3-the SC Pump Room Return to Outside 13,20,28,40 (050-A04A) 41.40 4.28 2.95 (Comp. Bldg)(3)

Sum 9.25 Access from Outside 43.33 5.82 4.21 (Comp. Bldg)(1)

Access Area outside Operation 5.90 30 2.95 Fig. 12.3-the SC Pump Room Return to Outside 13,20,28 (050-A04B) 48.15 3.43 2.75 (Aux. Bldg)(2)

Sum 9.91 (1) Section within the row line from PE to PF along the column line 31 in Figure 12.3-13 (2) Section within the column line from 23 to 24 along the row line AK in Figure 12.3-28 (3) Section within the row line from PI to PJ along the column line 31 in Figure 12.3-13 12.4-36 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (10 of 21)

e. Vital area mission dose for areas requiring infrequent access to perform post-accident vital functions (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after LOCA) (1 of 4)

Mission Dose Vital Area Access Action Dose Rate Occupancy Dose (Room No.) Route (mSv/hr) Time (min) (mSv)

Post-accident Sample Access from Outside 43.18 6.07 4.37 and Control Panel (Comp. Bldg)(1) Fig. 12.3-Room Sampling 7.11 20 2.37 13,20,28 (055-A48B, A49B) Sum 6.74 Access from Post-accident Sample and 40.91 6.47 4.41 Radiochemistry Lab Control Panel Room Fig. 12.3-(085-P37) Analysis 6.58 60 6.58 12,13,20,28 Return to MCR/TSC 6.73 6.45 0.72 Sum 11.72 Access from Post-accident Sample and 39.30 6.79 4.45 Sample Counting Control Panel Room Fig. 12.3-Room Analysis 6.58 60 6.58 12,13,20,28 (085-P36)

Return to MCR/TSC 6.73 6.77 0.76 Sum 11.79 Access from Outside 6.73 5.78 0.65 (Comp. Bldg)(1)

Remote Shutdown Operation 6.58 90 9.88 Fig. 12.3-Room Return to Outside 13,28,36 (137-A06D) 6.73 3.35 0.38 (Aux. Bldg)(2)

Sum 10.90 Access from MCR/TSC 7.74 6.12 0.79 Remote Control Operation 6.58 90 9.88 Fig. 12.3-Console Room Return to MCR/TSC 7.74 6.12 0.79 32,36,40 (137-A41A)

Sum 11.45 (1) Section within the row line from PE to PF along the column line 31 in Figure 12.3-13 (2) Section within the column line from 14 to 15 along the row line AK in Figure 12.3-28 12.4-37 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (11 of 21)

e. Vital area mission dose for areas requiring infrequent access to perform post-accident vital functions (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after LOCA) (2 of 4)

Mission Dose Vital Area Access Action Dose Rate Occupancy Dose (Room No.) Route (mSv/hr) Time (min) (mSv)

Access from 7.58 6.35 0.80 Class 1E Switchgear MCR/TSC Fig. 12.3-01A Room Inspection 142.18 10 23.70 24,40 (078-A25A) Return to MCR/TSC 7.58 6.35 0.80 Sum 25.30 Access from 7.58 6.31 0.80 Class 1E Switchgear MCR/TSC Fig. 12.3-01B Room Inspection 145.30 10 24.22 24,40 (078-A25B) Return to MCR/TSC 7.58 6.31 0.80 Sum 25.81 Access from 7.71 5.30 0.68 Class 1E Switchgear MCR/TSC Fig. 12.3-01C Room Inspection 6.58 10 1.10 24,40 (078-A02C) Return to MCR/TSC 7.71 5.30 0.68 Sum 2.46 Access from 7.71 5.26 0.68 Class 1E Switchgear MCR/TSC Inspection 6.58 10 1.10 Fig. 12.3-01D Room Return to MCR/TSC 7.71 5.26 0.68 24,40 (078-A02D)

Sum 2.45 Access from Outside 6.73 5.38 0.60 (Comp. Bldg)(1)

I&C Equip. Room Inspection 24.07 30 12.04 Fig. 12.3-(157-A19C) Return to Outside 13,28,40 6.73 4.10 0.46 (Aux. Bldg)(2)

Sum 13.10 (1) Section within the row line from PE to PF along the column line 31 in Figure 12.3-13 (2) Section within the column line from 14 to 15 along the row line AA in Figure 12.3-28 12.4-38 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (12 of 21)

e. Vital area mission dose for areas requiring infrequent access to perform post-accident vital functions (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after LOCA) (3 of 4)

Mission Dose Vital Area Access Action Dose Rate Occupancy Dose (Room No.) Route (mSv/hr) Time (min) (mSv)

Access from Outside 6.73 5.31 0.60 (Comp. Bldg)(1)

I&C Equip. Room Inspection 6.58 30 3.29 Fig. 12.3-(157-A25C) Return to Outside 13,28,40 6.73 4.04 0.45 (Aux. Bldg)(2)

Sum 4.34 Access from Outside 6.73 6.29 0.71 (Comp. Bldg)(1)

I&C Equip. Room Inspection 6.58 30 3.29 Fig. 12.3-(157-A01D) Return to Outside 13,28,40 6.73 3.87 0.43 (Aux. Bldg)(3)

Sum 4.43 Access from Outside 6.73 6.52 0.73 (Comp. Bldg)(1)

I&C Equip. Room Inspection 24.07 30 12.04 Fig. 12.3-(157-A19D) Return to Outside 13,28,40 6.73 4.09 0.46 (Aux. Bldg)(3)

Sum 13.23 Access from 23.30 9.41 3.65 MCR/TSC Access Area outside Operation 7.41 30 3.70 Fig. 12.3-the CS Pump Room Return to Outside 13,20,28,40 (050-A01C) 38.38 4.93 3.15 (Comp. Bldg)(4)

Sum 10.51 (1) Section within the row line from PE to PF along the column line 31 in Figure 12.3-13 (2) Section within the column line from 14 to 15 along the row line AA in Figure 12.3-28 (3) Section within the column line from 14 to 15 along the row line AK in Figure 12.3-28 (4) Section within the row line from PI to PJ along the column line 31 in Figure 12.3-13 12.4-39 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (13 of 21)

e. Vital area mission dose for areas requiring infrequent access to perform post-accident vital functions (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after LOCA) (4 of 4)

Mission Dose Vital Area Access Action Dose Rate Occupancy Dose (Room No.) Route (mSv/hr) Time (min) (mSv)

Access from Outside 41.20 6.47 4.44 (Comp. Bldg)(1)

Access Area outside Operation 7.41 30 3.70 Fig. 12.3-the CS Pump Room Return to Outside 13,20,28 (050-A01D) 42.76 4.11 2.93 (Aux. Bldg)(2)

Sum 11.08 Access from 24.39 8.77 3.56 MCR/TSC Access Area outside Operation 7.41 30 3.70 Fig. 12.3-the SC Pump Room Return to Outside 13,20,28,40 (050-A04A) 42.90 4.28 3.06 (Comp. Bldg)(3)

Sum 10.33 Access from Outside 44.84 5.82 4.35 (Comp. Bldg)(1)

Access Area outside Operation 7.41 30 3.70 Fig. 12.3-the SC Pump Room Return to Outside 13,20,28 (050-A04B) 49.65 3.43 2.84 (Aux. Bldg)(2)

Sum 10.89 (1) Section within the row line from PE to PF along the column line 31 in Figure 12.3-13 (2) Section within the column line from 23 to 24 along the row line AK in Figure 12.3-28 (3) Section within the row line from PI to PJ along the column line 31 in Figure 12.3-13 12.4-40 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (14 of 21)

f. Vital area mission dose for areas requiring infrequent access to perform post-accident vital functions (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after LOCA) (1 of 4)

Mission Dose Vital Area Access Action Dose Rate Occupancy Dose (Room No.) Route (mSv/hr) Time (min) (mSv)

Post-accident Sample Access from Outside 40.54 6.07 4.10 and Control Panel (Comp. Bldg)(1) Fig. 12.3-Room Sampling 4.47 20 1.49 13,20,28 (055-A48B, A49B) Sum 5.59 Access from Post-accident Sample and 38.27 6.47 4.13 Radiochemistry Lab Control Panel Room Fig. 12.3-(085-P37) Analysis 3.94 60 3.94 12,13,20,28 Return to MCR/TSC 4.09 6.45 0.44 Sum 8.51 Access from Post-accident Sample and 36.66 6.79 4.15 Sample Counting Control Panel Room Fig. 12.3-Room Analysis 3.94 60 3.94 12,13,20,28 (085-P36)

Return to MCR/TSC 4.09 6.77 0.46 Sum 8.56 Access from Outside 4.09 5.78 0.39 (Comp. Bldg)(1)

Remote Shutdown Operation 3.94 90 5.92 Fig. 12.3-Room Return to Outside 13,28,36 (137-A06D) 4.09 3.35 0.23 (Aux. Bldg)(2)

Sum 6.54 Access from MCR/TSC 5.09 6.12 0.52 Remote Control Operation 3.94 90 5.92 Fig. 12.3-Console Room Return to MCR/TSC 5.09 6.12 0.52 32,36,40 (137-A41A)

Sum 6.96 (1) Section within the row line from PE to PF along the column line 31 in Figure 12.3-13 (2) Section within the column line from 14 to 15 along the row line AK in Figure 12.3-28 12.4-41 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (15 of 21)

f. Vital area mission dose for areas requiring infrequent access to perform post-accident vital functions (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after LOCA) (2 of 4)

Mission Dose Vital Area Access Action Dose Rate Occupancy Dose (Room No.) Route (mSv/hr) Time (min) (mSv)

Access from MCR/TSC 4.94 6.35 0.52 Class 1E Switchgear Inspection 139.54 10 23.26 Fig. 12.3-01A Room Return to Outside 13.42 2.78 0.62 13,24,28,40 (078-A25A) (Comp. Bldg)(1)

Sum 24.40 Access from MCR/TSC 4.94 6.31 0.52 Class 1E Switchgear Inspection 142.66 10 23.78 Fig. 12.3-01B Room Return to Outside 16.22 2.03 0.55 24,28,40 (078-A25B) (Aux. Bldg)(2)

Sum 24.85 Access from MCR/TSC 5.07 5.30 0.45 Class 1E Switchgear Inspection 3.94 10 0.66 Fig. 12.3-01C Room Return to MCR/TSC 5.07 5.30 0.45 24,40 (078-A02C)

Sum 1.55 Access from MCR/TSC 5.07 5.26 0.45 Class 1E Switchgear Inspection 3.94 10 0.66 Fig. 12.3-01D Room Return to MCR/TSC 5.07 5.26 0.45 24,40 (078-A02D)

Sum 1.55 Access from Outside 4.09 5.38 0.37 (Comp. Bldg)(3)

I&C Equip. Room Inspection 21.43 30 10.72 Fig. 12.3-(157-A19C) Return to Outside 13,28,40 4.09 4.10 0.28 (Aux. Bldg)(4)

Sum 11.36 (1) Section within the row line from PI to PJ along the column line 31 in Figure 12.3-13 (2) Section within the column line from 23 to 24 along the row line AK in Figure 12.3-28 (3) Section within the row line from PE to PF along the column line 31 in Figure 12.3-13 (4) Section within the column line from 14 to 15 along the row line AA in Figure 12.3-28 12.4-42 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (16 of 21)

f. Vital area mission dose for areas requiring infrequent access to perform post-accident vital functions (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after LOCA) (3 of 4)

Mission Dose Vital Area Access Action Dose Rate Occupancy Dose (Room No.) Route (mSv/hr) Time (min) (mSv)

Access from Outside 4.09 5.31 0.36 (Comp. Bldg)(1)

I&C Equip. Room Inspection 3.94 30 1.97 Fig. 12.3-(157-A25C) Return to Outside 13,28,40 4.09 4.04 0.28 (Aux. Bldg)(2)

Sum 2.61 Access from Outside 4.09 6.29 0.43 (Comp. Bldg)(1)

I&C Equip. Room Inspection 3.94 30 1.97 Fig. 12.3-(157-A01D) Return to Outside 13,28,40 4.09 3.87 0.26 (Aux. Bldg)(3)

Sum 2.67 Access from Outside 4.09 6.52 0.44 (Comp. Bldg)(1)

I&C Equip. Room Inspection 21.43 30 10.72 Fig. 12.3-(157-A19D) Return to Outside 13,28,40 4.09 4.09 0.28 (Aux. Bldg)(3)

Sum 11.44 Access from 20.65 9.41 3.24 MCR/TSC Access Area outside Operation 4.77 30 2.38 Fig. 12.3-the CS Pump Room Return to Outside 13,20,28,40 (050-A01C) 35.74 4.93 2.93 (Comp. Bldg)(4)

Sum 8.56 (1) Section within the row line from PE to PF along the column line 31 in Figure 12.3-13 (2) Section within the column line from 14 to 15 along the row line AA in Figure 12.3-28 (3) Section within the column line from 14 to 15 along the row line AK in Figure 12.3-28 (4) Section within the row line from PI to PJ along the column line 31 in Figure 12.3-13 12.4-43 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (17 of 21)

f. Vital area mission dose for areas requiring infrequent access to perform post-accident vital functions (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after LOCA) (4 of 4)

Mission Dose Vital Area Access Action Dose Rate Occupancy Dose (Room No.) Route (mSv/hr) Time (min) (mSv)

Access from Outside 38.56 6.47 4.16 (Comp. Bldg)(1)

Access Area outside Operation 4.77 30 2.38 Fig. 12.3-the CS Pump Room Return to Outside 13,20,28 (050-A01D) 40.12 4.11 2.75 (Aux. Bldg)(2)

Sum 9.29 Access from 21.75 8.77 3.18 MCR/TSC Access Area outside Operation 4.77 30 2.38 Fig. 12.3-the SC Pump Room Return to Outside 13,20,28,40 (050-A04A) 40.26 4.28 2.87 (Comp. Bldg)(3)

Sum 8.44 Access from Outside 42.20 5.82 4.10 (Comp. Bldg)(1)

Access Area outside Operation 4.77 30 2.38 Fig. 12.3-the SC Pump Room Return to Outside 13,20,28 (050-A04B) 47.01 3.43 2.68 (Aux. Bldg)(2)

Sum 9.17 (1) Section within the row line from PE to PF along the column line 31 in Figure 12.3-13 (2) Section within the column line from 23 to 24 along the row line AK in Figure 12.3-28 (3) Section within the row line from PI to PJ along the column line 31 in Figure 12.3-13 12.4-44 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (18 of 21)

g. Vital area mission dose for areas requiring infrequent access to perform post-accident vital functions (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after LOCA) (1 of 4)

Mission Dose Vital Area Access Action Dose Rate Occupancy Dose (Room No.) Route (mSv/hr) Time (min) (mSv)

Post-accident Sample Access from Outside 4.21 6.07 0.43 and Control Panel (Comp. Bldg)(1) Fig. 12.3-Room Sampling 0.77 20 0.26 13,21,29 (055-A48B, A49B) Sum 0.68 Access from Post-accident Sample and 4.01 6.47 0.43 Radiochemistry Lab Control Panel Room Fig. 12.3-(085-P37) Analysis 0.77 60 0.77 12,13,21,29 Return to MCR/TSC 0.92 6.45 0.10 Sum 1.30 Access from Post-accident Sample and 3.86 6.79 0.44 Sample Counting Control Panel Room Fig. 12.3-Room Analysis 0.77 60 0.77 12,13,21,29 (085-P36)

Return to MCR/TSC 0.92 6.77 0.10 Sum 1.31 Access from Outside 0.92 5.78 0.09 (Comp. Bldg)(1)

Remote Shutdown Operation 0.77 90 1.15 Fig. 12.3-Room Return to Outside 13,29,37 (137-A06D) 0.92 3.35 0.05 (Aux. Bldg)(2)

Sum 1.29 Access from 1.92 6.12 0.20 Remote Control MCR/TSC Fig. 12.3-Console Room Operation 0.77 90 1.15 33,37,41 (137-A41A) Return to MCR/TSC 1.92 6.12 0.20 Sum 1.54 (1) Section within the row line from PE to PF along the column line 31 in Figure 12.3-13 (2) Section within the column line from 14 to 15 along the row line AK in Figure 12.3-29 12.4-45 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (19 of 21)

g. Vital area mission dose for areas requiring infrequent access to perform post-accident vital functions (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after LOCA) (2 of 4)

Mission Dose Vital Area Access Action Dose Rate Occupancy Dose (Room No.) Route (mSv/hr) Time (min) (mSv)

Access from 0.99 6.35 0.10 Class 1E Switchgear MCR/TSC Fig. 12.3-01A Room Inspection 10.99 10 1.83 25,41 (078-A25A) Return to MCR/TSC 0.99 6.35 0.10 Sum 2.04 Access from 0.99 6.31 0.10 Class 1E Switchgear MCR/TSC Fig. 12.3-01B Room Inspection 11.22 10 1.87 25,41 (078-A25B) Return to MCR/TSC 0.99 6.31 0.10 Sum 2.08 Access from 1.00 5.30 0.09 Class 1E Switchgear MCR/TSC Fig. 12.3-01C Room Inspection 0.77 10 0.13 25,41 (078-A02C) Return to MCR/TSC 1.00 5.30 0.09 Sum 0.31 Access from 1.00 5.26 0.09 Class 1E Switchgear MCR/TSC Inspection 0.77 10 0.13 Fig. 12.3-01D Room Return to MCR/TSC 1.00 5.26 0.09 25,41 (078-A02D)

Sum 0.30 Access from Outside 0.92 5.38 0.08 (Comp. Bldg)(1)

I&C Equip. Room Inspection 0.87 30 0.44 Fig. 12.3-(157-A19C) Return to Outside 13,29,41 0.92 4.10 0.06 (Aux. Bldg)(2)

Sum 0.58 (1) Section within the row line from PE to PF along the column line 31 in Figure 12.3-13 (2) Section within the column line from 14 to 15 along the row line AA in Figure 12.3-29 12.4-46 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (20 of 21)

g. Vital area mission dose for areas requiring infrequent access to perform post-accident vital functions (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after LOCA) (3 of 4)

Mission Dose Vital Area Access Action Dose Rate Occupancy Dose (Room No.) Route (mSv/hr) Time (min) (mSv)

Access from Outside 0.92 5.31 0.08 (Comp. Bldg)(1)

I&C Equip. Room Inspection 0.77 30 0.38 Fig. 12.3-(157-A25C) Return to Outside 13,29,41 0.92 4.04 0.06 (Aux. Bldg)(2)

Sum 0.53 Access from Outside 0.92 6.29 0.10 (Comp. Bldg)(1)

I&C Equip. Room Inspection 0.77 30 0.38 Fig. 12.3-(157-A01D) Return to Outside 13,29,41 0.92 3.87 0.06 (Aux. Bldg)(3)

Sum 0.54 Access from Outside 0.92 6.52 0.10 (Comp. Bldg)(1)

I&C Equip. Room Inspection 0.87 30 0.44 Fig. 12.3-(157-A19D) Return to Outside 13,29,41 0.92 4.09 0.06 (Aux. Bldg)(3)

Sum 0.60 Access from 2.33 9.41 0.36 MCR/TSC Access Area outside Operation 0.77 30 0.38 Fig. 12.3-the CS Pump Room Return to Outside 13,21,29,41 (050-A01C) 3.61 4.93 0.30 (Comp. Bldg)(4)

Sum 1.04 (1) Section within the row line from PE to PF along the column line 31 in Figure 12.3-13 (2) Section within the column line from 14 to 15 along the row line AA in Figure 12.3-29 (3) Section within the column line from 14 to 15 along the row line AK in Figure 12.3-29 (4) Section within the row line from PI to PJ along the column line 31 in Figure 12.3-13 12.4-47 Rev. 2

APR1400 DCD TIER 2 Table 12.4-8 (21 of 21)

g. Vital area mission dose for areas requiring infrequent access to perform post-accident vital functions (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after LOCA) (4 of 4)

Mission Dose Vital Area Access Action Dose Rate Occupancy Dose (Room No.) Route (mSv/hr) Time (min) (mSv)

Access from Outside 4.01 6.47 0.43 (Comp. Bldg)(1)

Access Area outside Operation 0.77 30 0.38 Fig. 12.3-the CS Pump Room Return to Outside 13,21,29 (050-A01D) 3.98 4.11 0.27 (Aux. Bldg)(2)

Sum 1.09 Access from 2.43 8.77 0.35 MCR/TSC Access Area outside Operation 0.77 30 0.38 Fig. 12.3-the SC Pump Room Return to Outside 13,21,29,41 (050-A04A) 4.01 4.28 0.29 (Comp. Bldg)(3)

Sum 1.02 Access from Outside 4.35 5.82 0.42 (Comp. Bldg)(1)

Access Area outside Operation 0.77 30 0.38 Fig. 12.3-the SC Pump Room Return to Outside 13,21,29 (050-A04B) 4.59 3.43 0.26 (Aux. Bldg)(2)

Sum 1.07 (1) Section within the row line from PE to PF along the column line 31 in Figure 12.3-13 (2) Section within the column line from 23 to 24 along the row line AK in Figure 12.3-29 (3) Section within the row line from PI to PJ along the column line 31 in Figure 12.3-13 12.4-48 Rev. 2

APR1400 DCD TIER 2 Table 12.4-9 (1 of 2)

APR1400 Plant Systems Requiring Full NRC RG 4.21 Evaluations System Sheet No.

Code System Title In Table 12.4-10 AF Auxiliary Feedwater 50 AS Auxiliary Steam 63 AT Auxiliary Feedwater Pump Turbine 50 CA Condenser Vacuum 55 CC Component Cooling Water 24 CD Condensate 47 CM Containment Monitoring 73 CP Condensate Polishing 52 CS Containment Spray 12 CV Chemical and Volume Control 18 CW Circulating Water 57 DE Radioactive Drain 39 DM Miscellaneous Building Drain 69 DT Turbine Generator Building Drain 69 ES Extraction Steam 47 FC Spent Fuel Pool Cooling and Cleanup 30 FT Feedwater Pump Turbine 47 FW Feedwater 47 GW Gaseous Waste Management 32 HD Heater Drain 47 HG Containment Hydrogen Control 14 IW In-Containment Water Storage 15 12.4-49 Rev. 2

APR1400 DCD TIER 2 Table 12.4-9 (2 of 2)

System Sheet No.

Code System Title In Table 12.4-10 MS Main Steam 45 PR Radiation Monitoring 71 PS Process Sampling 42 PX Primary Sampling 42 RC Reactor Coolant 1 RG Reactor Coolant Gas Vent 7 SC Shutdown Cooling 10 SD Steam Generator Blowdown 22 SI Safety Injection 8 SX Essential Service Water 28 TA Main Turbine and Auxiliaries 45 VB Compound Building HVAC 66 VF Fuel Handling Area HVAC 66 VK Auxiliary Building Controlled Area HVAC 66 VP Reactor Containment Building HVAC 66 VQ Reactor Containment Building Purge 66 WH Turbine Generator Building Open Cooling Water 59 WT Turbine Generator Building Closed Cooling Water 61 WV Liquid Waste Management 34 WX Solid Waste Management 36 12.4-50 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (1 of 75)

NRC RG 4.21 Design Objectives and Applicable DCD Subsection Information for Minimizing Contamination and Generation of Radioactive Waste System: Reactor Coolant System DCD Tier 2 Objective SSC Control Measures(1) / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The RCS is designed in accordance with ASME Section III and proven nuclear power industry 5.4.3 of Unintended experience. The RCS is designed with an over-pressure protection system and other Contamination engineered safety features (ESF) to provide reasonable assurance of safe operation and mitigation of accident conditions, thus preventing unintended contamination.

• The RCS components are designed to provide the first barrier to contain the radioactivity resulting from power operation and damaged fuel. These components are designed with proven materials and fabrication techniques, and are housed inside the reactor containment structure, which provides the second barrier. The basemat of the reactor containment structure is equipped with steel liners to prevent and minimize the spread of contamination from leaked reactor coolant. The in-core instrumentation (ICI) sump is provided at the basemat level to collect leaks, and the reactor containment sump is provided at the reactor vessel operating floor to collect leaks from the other components. The reactor floor is also lined with a steel layer to prevent contamination of the concrete structure.

• The system design, including wetted parts of the reactor vessel, the steam generators, the pressurizer, the reactor coolant pumps, and the associated piping, is designed with stainless steel cladding or base material, and has welded construction for life-cycle planning, thus minimizing leakage and unintended contamination of the facility and the environment.

• The components are designed with material resistant to primary water stress corrosion cracking and reduced SG hot leg temperature to minimize the potential for stress corrosion and cracking.

• The containment also includes the IRWST and a holdup volume tank with sufficient capacity to provide temporary storage of reactor coolant from anticipated operational occurrences and design basis accidents.

12.4-51 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (2 of 75)

DCD Tier 2 Objective SSC Control Measures(1) / Design Features in DCD to Meet Objective Reference 2 Adequate and Early Leak

• The RCS components are designed with multiple level, pressure, and temperature, and radiation 5.4.3 Detection detection instruments to provide reasonable assurance of safe operation of the SSCs, including the associated piping, and provide alarms to the operating personnel in the event of off-normal conditions. Releases from over-pressurization and leaks are collected in the reactor drain tank (RDT), the ICI sump, the containment sump, and the IRWST. A detailed description of the early leak detection instruments, locations, leak collections and pathways, and operation notifications for response actions is presented in Subsection 5.2.5.

• Leak detection methods are segregated into two classifications: unidentified leakages and Identified leakages. Unidentified leakages are monitored by containment sump level, area particulate radiation detection, and containment atmosphere humidity instruments. Leakages are collected in sumps and a computer program is developed to estimate leakage rates based on flow measurements. This approach can detect 0.5 gpm leakage within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Additionally, 19 locations within the containments are equipped with acoustic leak detection instruments where leakage is considered critical. See Subsection 5.2.5.1.1 for details.

• Containment area radiation monitors, temperature and pressure monitors, sump levels and sump pump flow rates are programmed to provide leakage information and notification in the MCR for operator actions.

• Acoustic leak monitoring instruments are installed in other critical areas within the reactor coolant pressure boundary, including in-core instrumentation nozzles at reactor lower head, SG man-way seals, reactor hot and cold leg welding areas, reactor coolant pumps shaft seal housings, pressurizer lower head area, and CEA nozzle areas near the reactor head. These leak monitoring instruments provide early detection of leak and alarm signals for operator actions when required.

• The identified leakages for the primary RCS leak detection methods and locations, and other indications of reactor coolant leakage are summarized in Subsections 5.2.5.1.2. The leakage sources and instruments for early detection of reactor coolant leakage instruments are as follows:

include SG tube and tubesheet leaks, pressurizer pilot operated safety relief valve, reactor coolant pump seals, valves, reactor vessel head flange seal, and SG tube and tubesheet leakages.

12.4-52 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (3 of 75)

DCD Tier 2 Objective SSC Control Measures(1) / Design Features in DCD to Meet Objective Reference 2 (cont.)

• Steam generator leaks are monitored by N-16 monitors on each of the main steam lines. An increase in radioactivity, as indicated by the condenser vacuum vent effluent monitor, and SG blowdown monitor will reveal reactor coolant leakage through SG tubes to the secondary side.

Routine analysis of the SG secondary side will also indicate leakage of reactor coolant into the secondary system.

• The RCPs are designed with two mechanical seals and one mechanical vapor seal made of renewable silicon carbide rings. Each of these three seals can withstand full operating pressure.

Any leakage from these seals is collected and routed to the RDT. An acoustic sensor is provided with each pump to detect any leak through each of the seal.

• Temperatures downstream of the pilot-operated safety relief valves (POSRVs) are continuously monitored for early detection of leaks.

• Leakage between the two metal O-rings that seal the reactor vessel head flange is routed through a leakoff line to the RDT. A normally closed, remotely activated isolation valve and a pressure indicator are installed in the leakoff line. Leakage from the reactor vessel head flange causes pressure in the leakoff line to rise. The pressure in the leakoff line is continuously monitored to detect the presence of a leak. Any leakage is bled off to the RDT by opening the isolation valve.

• Other primary coolant leak detection methods and locations are discussed in Subsection 5.2.5.2.2.

3 Reduction of Cross-

• Each RCP is equipped with the oil collection system in accordance with NRC RG 1.189 that 5.4.3 Contamination, collect oil from non-welded parts. Leaked oil is collected and drained to an oil collection tank Decontamination, and that can store the entire inventory. This design approach prevents the potential generation of Waste Generation mixed waste.

• Shop welding of major piping to larger pieces is maximized to reduce the field welding. Piping safe-ends are provided for field welding connections to minimize the potential for leaking.

• The SSCs are designed with life-cycle planning through the use of nuclear-industry proven materials compatible with the chemical, physical, and radioactive environment, thus minimizing waste generation.

12.4-53 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (4 of 75)

DCD Tier 2 Objective SSC Control Measures(1) / Design Features in DCD to Meet Objective Reference 3 (cont.)

• The RCS is designed with continuous letdown of reactor coolant for purification in the CVCS system and provisions of sampling to maintain reactor coolant chemistry and quality. Similarly, the steam generators are designed with continuous blowdown of dissolved and suspended solids on the secondary side for treatment and with provisions for sampling and analysis. This design approach minimizes the buildup of radioactivity and impurities, enhance safe operation, and reduces the decontamination and waste generation during decommissioning.

• The RCS components are made of steel with low cobalt content as much as it is practicable. The approach minimizes waste generation.

• Manual valves 2 inches and larger are designed with a double-packed stem and a lantern ring with an intermediate leak drain connection. The drains are routed to the drain system for collection and treatment. This design approach minimizes the spread of contamination and the contamination of the facility.

• Sampling lines from the RCS are designed to be as short and straight as possible to minimize traps and pockets. Flow velocity is maintained in the turbulent region to prevent settling of suspended solids. Gaseous sampling lines are sloped upward to facilitate draining of condensate. This design approach minimizes accumulation of condensate and provides more accurate samples, and thus minimizing waste generation. Sampling lines are purged with adequate demineralized water or nitrogen gas in order to maintain quality of sampling.

• The process piping containing primary coolant is properly designed to be the shortest distance between components, sized for turbulent flow, and laid out to minimize traps and erosion. This approach facilitates easier flow and with sufficient velocities to prevent settling of solids, thus reducing the decontamination needs and waste generation.

• Utility connections are designed with a minimum of two barriers to prevent cross-contamination.

12.4-54 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (5 of 75)

DCD Tier 2 Objective SSC Control Measures(1) / Design Features in DCD to Meet Objective Reference 4 Decommissioning

• The top surface of the containment basemat is sloped and lined with a layer of steel to prevent 5.4.3 Planning contamination of the basemat by facilitating draining of leakage to the ICI sump and the containment sump. This layered approach facilitates easier decommissioning and minimize waste generation.

• The SSCs are designed for the full service life and are fabricated as individual assemblies for easy removal.

• The RCS is designed with no embedded or buried piping, thus facilitating decommissioning planning and preventing unintended contamination.

5 Operations and

• The RCS is designed for remote and automatic operation with operator supervision. Adequate 5.4.3 Documentation instrumentation, including level, flow, temperature, and pressure elements, radiation monitors, and acoustic leak detection instruments, is provided to monitor and control the operations to prevent undue interruption, and minimize the spread of contamination and waste generation.

• The RCS is designed for a minimal amount of maintenance. The layout takes into account the necessary access, laydown areas, handling and lifting cranes, and the auxiliary equipment in the event that maintenance is required.

• The RCS components are equipped with loose parts monitoring instruments and vibration monitoring instruments at specific locations. This design approach enhances equipment life and minimizes waste generation.

• Areas adjacent to the RCS pressure boundary are designed to have adequate clearance to access for inservice inspections. Personnel access is provided for component maintenance and inspections.

• The COL applicant is to prepare operational procedures and maintenance programs for leak detection and contamination control of the RCS (COL 5.4(1)). Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

• Complete documentation of design, construction, design modifications, field changes, and operations is to be maintained by the COL applicant. Documentation requirements are included as a COL information item (COL 5.4(2)).

12.4-55 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (6 of 75)

DCD Tier 2 Objective SSC Control Measures(1) / Design Features in DCD to Meet Objective Reference 6 Site Radiological

• The RCS is located inside the reactor containment building. Because of its location and 5.4.3 Environmental associated safety design features, the potential for environmental contamination of soil and Monitoring groundwater from reactor coolant is minimal. Therefore, RCS inclusion in the site radiological environmental monitoring program is not required. However, a site radiological environmental monitoring program is included for the whole plant for detection radiological of contamination.

(1) The SSC Control Measures consist of excerpts/summary from the referenced DCD sections.

12.4-56 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (7 of 75)

System: Reactor Coolant Gas Vent System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The RCGVS is designed to vent non-condensable gases from the pressurizer and the reactor 5.4.12 of Unintended vessel upper head and depressurizes the reactor coolant system in the event that the pressurizer Contamination main spray or auxiliary spray systems are unavailable during plant cooldown. The piping directs the vented gases to the IRWST or the reactor drain tank (RDT) and is sloped to facilitate the drainage of condensation, thus minimizing leakage and unintended contamination of the facility and the environment.

2 Adequate and Early Leak

• The RCGVS is designed not to be used during normal operation, and the piping is designed to 5.4.12 Detection slope downward to drain to the RDT and the IRWST. The potential for leakage is very low and a leak detection system is not required.

3 Reduction of Cross-

• The RCGVS piping is made of stainless steel material for life-cycle planning. The material is 5.4.12 Contamination, compatible with the chemical, physical, and radiological environment, thus minimizing waste Decontamination, and generation.

Waste Generation 4 Decommissioning

• The RCGVS piping is designed for the full service life of the plant and is accessible for easy 5.4.12 Planning removal during decommissioning.

5 Operations and

• The RCGVS piping is used during refueling or post-accident conditions. Adequate 5.4.12 Documentation instrumentation is provided to indicate the venting operations to prevent undue interruption.

6 Site Radiological

• The RCGVS piping is part of the overall plant and is designed to vent non-condensable gases 5.4.12 Environmental from the reactor vessel and the pressurizer. The RCGVS piping does not generate any Monitoring radioactive materials and thus, site radiological environmental monitoring is not required.

12.4-57 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (8 of 75)

System: Safety Injection System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The SIS components are located inside the reactor containment building and the auxiliary 6.3.6 of Unintended building. The cubicle floors, where the components are located, are sloped, coated with epoxy, Contamination and provided with drains that are routed to the local drain hubs and sumps. This design approach prevents unintended contamination of the facility and the environment.

• The system tanks and pumps are fabricated from stainless steel material and are of welded construction for life-cycle planning, thus minimizing leakage and unintended contamination of the facility and the environment.

2 Adequate and Early Leak

• The SI tanks are equipped with level instruments to monitor liquid levels. The level instruments 6.3.6 Detection provide alarm signals in the event that leakage occurs.

• The SIS is not used during normal power operation. Any leakage is drained to the floor and is collected in the local sump which is equipped with a liquid level switch. If leakage exceeds a predetermined liquid level within the sump, the level switch initiates an alarm in the MCR for operator actions to investigate the source of leakage.

3 Reduction of Cross-

• The SSCs are designed with life-cycle planning using nuclear industry-proven materials 6.3.6 Contamination, compatible with the chemical, physical, and radiological environment, thus minimizing waste Decontamination, and generation. The pump shaft seals are of the mechanical type and constructed of materials Waste Generation compatible with the SI fluid. This design approach minimizes contamination of the facility and the environment. The SIS system is designed to be isolated from the RCS during normal power operation to minimize cross-contamination of systems.

4 Decommissioning

• Design features such as welding techniques used and surface finishes are intended to minimize 6.3.6 Planning the need for decontamination and the resultant waste generation. The SIS is designed with a minimal amount of embedded piping for contaminated or potentially contaminated fluid, to facilitate decommissioning.

5 Operations and

• The COL applicant is to prepare operational procedures and maintenance programs as related to 6.3.6 Documentation leak detection and contamination control (COL 6.3(1)). Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

• The COL applicant is to maintain complete documentation of system design, construction, design modifications, field changes, and operations (COL 6.3(2)).

12.4-58 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (9 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 6 Site Radiological

• The SIS is on standby mode during normal power operation and is designed to have low levels of 6.3.6 Environmental Monitoring contamination. Through monitoring, in-service inspection, and lessons learned from industry experiences, the integrity of the SIS is well maintained, resulting in a very low potential for contamination of the facility. Because of its location, the potential for environmental contamination of soil and groundwater from liquid leakage is minimal. Hence, the SIS is not required to be part of the site radiological environmental monitoring program.

12.4-59 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (10 of 75)

System: Shutdown Cooling System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The SCS components are located in individual cubicles inside the auxiliary building. The floors 5.4.7 of Unintended are sloped, coated with epoxy, and provided with drains that are routed to the local drain hubs and Contamination sumps. This design approach prevents unintended contamination of the facility and the environment.

• The SCS components (pumps, heat exchangers, piping) are fabricated from stainless steel material and are of welded construction, thus minimizing leakage and unintended contamination of the facility and the environment.

2 Adequate and Early Leak

• The SCS is used only in post-shutdown periods. Any leakage is drained to the floor and is 5.4.7 Detection collected in the local sump, which is equipped with a liquid level switch. If leakage exceeds a predetermined liquid level within the sump, the level switch initiates an alarm in the MCR for operator actions to investigate the source of leakage.

3 Reduction of Cross-

• The SSCs are designed with life-cycle planning through the use of nuclear industry-proven 5.4.7 Contamination, materials compatible with the chemical, physical, and radiological environment, thus minimizing Decontamination, and waste generation.

Waste Generation

• The pumps are equipped with drains directly routed to radioactive drain system. This design approach minimizes contamination to the facility and the environment.

• The SCS is provided with isolation from the RCS when the RCS is at high pressure (during normal power operation). The interlocks associated with six valves on the two SCS suction lines are provided to prevent the valves from opening in the event RCS pressure exceeds SCS operating pressure, thus minimizing cross-contamination between the systems.

4 Decommissioning

• The SSCs are designed for extended service life and are fabricated as individual assemblies for 5.4.7 Planning easy removal.

• The SSCs are designed with decontamination capabilities. Design features, such as the welding techniques used, surface finishes, etc., are included to minimize the need for decontamination and the resultant waste generation.

• The SCS is designed with minimal embedded piping for contaminated or potentially contaminated fluid, which minimizes the potential for unintended contamination of the environment.

12.4-60 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (11 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 5 Operations and

• Operational procedures and maintenance programs as related to leak detection and contamination 5.4.7 Documentation control will be prepared by the COL applicant. Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

• Complete documentation of system design, construction, design modifications, field changes, and operations is to be maintained by the COL applicant. Documentation requirements are included as a COL information item.

6 Site Radiological

• The SCS is on standby mode during normal power operation and is designed to have low levels of 5.4.7 Environmental Monitoring contamination. Through monitoring, inservice inspection, and lessons learned from industry experiences, the integrity of the SCS is well maintained, resulting in a very low level of contamination of the facility. Hence, the SCS is not required to be part of the site radiological environmental monitoring program.

12.4-61 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (12 of 75)

System: Containment Spray System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The CSS components are located in individual cubicles inside the auxiliary building. The 6.5.2 of Unintended cubicle floors are sloped, coated with epoxy, and provided with drains that are routed to the local Contamination drain hubs. This design approach prevents unintended contamination of the facility and the environment.

• The system heat exchangers and pumps are fabricated with stainless steel material and are of welded construction for life-cycle planning, thus minimizing leakage and unintended contamination of the facility and the environment.

• The pump shaft seal is the mechanical type and is constructed of materials compatible with the spray fluids. A flow restrictor is provided to minimize the loss of fluid in the event of a seal failure. This design approach minimizes contamination to the facility and the environment.

2 Adequate and Early Leak

• The CSS is not used during normal power operation. Any leakage is drained to the floor and is 6.5.2 Detection collected in the local sump, which is equipped with a liquid level switch. If leakage exceeds a predetermined liquid level within the sump, the level switch initiates an alarm in the MCR for operator actions to investigate the source of leakage.

3 Reduction of Cross-

• The SSCs are designed with life-cycle planning through the use of nuclear industry-proven 6.5.2 Contamination, materials compatible with the chemical, physical, and radiological environment, thus minimizing Decontamination, and waste generation.

Waste Generation

• Shell and tube type heat exchangers are used to transfer heat from the CSS side to the component cooling water system side. The heat exchangers are designed with stainless steel tubes to minimize the potential for cross-contamination between containment spray and component cooling water. The leakage from the heat exchangers is collected in the local floor drain and sumps.

• Process sampling connections are provided to determine levels of contamination and determine treatment requirements.

12.4-62 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (13 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 4 Decommissioning

• The SSCs are designed for the full service life and fabricated as individual assemblies for easy 6.5.2 Planning removal.

• The SSCs are designed with decontamination capabilities. Design features (e.g., utilized welding technique, surface finishes) are included to minimize the need for decontamination and the resultant waste generation.

• The CSS is designed without any embedded or buried piping for contaminated or potentially contaminated fluid, which minimizes the potential for unintended contamination of the environment.

5 Operations and

• The COL applicant is to provide the operational procedures and maintenance program as related 6.5.2 Documentation to leak detection and contamination control (COL 6.5(1)).

• The COL applicant is to maintain the complete documentation of system design, construction, design modifications, field changes, and operations (COL 6.5(2)).

6 Site Radiological

• The CSS is on standby mode during normal power operation and designed to have only low levels 6.5.2 Environmental Monitoring of contamination. Through monitoring, in-service inspection, and lessons learned from industry experiences, the integrity of the CSS is well maintained, resulting in a very low level of contamination of the facility. Because of its location, the potential for environmental contamination of the soil and the groundwater from liquid leakage is minimal. Hence, the CSS is not required to be part of the Site Radiological Environmental Monitoring Program.

12.4-63 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (14 of 75)

System: Containment Hydrogen Control System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The passive autocatalytic recombiners (PARs) and hydrogen igniters (HIs) are designed to control 6.2.5 of Unintended the hydrogen by hydrogen recombination and hydrogen burning, respectively. There is no Contamination piping associated with the PARs and the HIs, and the system has no interfaces with non-contaminated systems, except electrical power supply, thus minimizing leakage and unintended contamination of the facility and the environment.

2 Adequate and Early Leak

• All components of the containment hydrogen control system are inside the containment or the 6.2.5 Detection IRWST and the components operate in atmospheric conditions. There is no leakage from the components and a leak detection system is not required.

3 Reduction of Cross-

• The PARs and the HIs are designed with life-cycle planning through the use of nuclear industry- 6.2.5 Contamination, proven materials compatible with the chemical, physical, and radiological environment, thus Decontamination, and minimizing waste generation.

Waste Generation

• The PARs and the HIs are located inside containment and the IRWST and are not in contact with contaminated liquid. Hence decontamination by water is not required.

4 Decommissioning

• The PARs and the HIs are modular units and are designed for the full service life of the plant. 6.2.5 Planning The PARs and the HIs are fabricated as individual assemblies for easy removal to the maximum extent practicable.

5 Operations and

• The PARs are self-actuated and the HIs are actuated manually from the MCR or RSR with 6.2.5 Documentation adequate instrumentation to the control the operation. This controls the hydrogen concentration inside the containment and minimizes the potential for fire and explosion hazards that could result in the spread of contamination and waste generation.

6 Site Radiological

• The containment hydrogen control system is part of the overall plant and is designed to prevent 6.2.5 Environmental Monitoring the buildup of hydrogen concentrations. The containment hydrogen control system does not generate any radioactive materials and is thus not required to be included in site radiological environmental monitoring.

12.4-64 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (15 of 75)

System: In-Containment Water Storage System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The IRWST components, primarily the IRWST and the HVT located inside containment, are 6.8.2 of Unintended constructed of reinforced concrete lined with austenitic stainless steel. The IRWST and the Contamination HVT are interconnected by two spillways. The bottom of the spillway is submerged below the surface of the water in the IRWST and the top of the spillway is below the IRWST ceiling during normal operation. This limits the amount of water vapor and possible gaseous radioactivity that can enter the containment atmosphere. This design approach prevents unintended contamination of the facility and the environment.

• The top of the basemat inside the containment is provided with a steel liner. This design approach prevents unintended contamination of the basemat and the environment.

• The IRWST uses the SFP cooling and cleanup system (FC) components to maintain and purify the water contents, thus minimizing the spread of contamination and unintended contamination of the facility and the environment.

• The system piping is fabricated from stainless steel to minimize corrosion and erosion.

Granulated tri-sodium phosphate (TSP) is provided in a stainless steel basket to minimize the corrosion of the stainless steel piping during LOCA conditions.

• The IRWST sump pits are designed with steel liner plates to provide impermeable barriers to prevent contamination of the basemat.

2 Adequate and Early Leak

• The IRWST is designed to prevent the spread of contamination due to overflow. Four wide 6.8.2 Detection range (0 to 100 percent) level indicating channels are provided locally and in the MCR.

Adequate instrumentation, including temperature and pressure elements, is provided to monitor and control the operations of preventing overpressurization or vacuum, thus minimizing the spread of contamination and waste generation

• The HVT has one narrow range and four wide range level indicating channels. The narrow range level channel is used to detect the presence of fluid in the HVT. Upon detection of water, the narrow range channel actuates an alarm in the MCR to alert for operator actions.

• The reactor cavity has four wide range level indicating channels. The four wide range channels provide level indication in the MCR to facilitate monitoring fluid level after an accident.

12.4-65 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (16 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 3 Reduction of Cross-

• The SSCs are designed with life-cycle planning through the use of materials compatible 6.8.2 Contamination, with the chemical, physical, and radiological environment, thus minimizing waste generation.

Decontamination, and

• During normal operation, the IRWST water can be cooled by the SC heat exchangers and Waste Generation purified by the FC system. The CVCS provides a means to add makeup and adjust the boron concentration in the IRWST. This design approach minimizes the spread of contamination to the facility and the environment, and waste generation.

4 Decommissioning

• The SSCs are designed for the full service life. All structures inside the IRWST are simple 6.8.2 Planning and smooth in contour to facilitate its lining with stainless steel. The bottom of the tank is flat to avoid having sharp corners. This design approach minimizes potential for damages to the liner, which can result in unintended leakage and contamination.

• The SSCs are designed with smooth surface finishes to minimize the need for decontamination and hence waste generation during decommissioning.

5 Operations and

• The IRWST is designed with adequate instrumentation for remote operation and monitoring. 6.8.2 Documentation

• A maintenance access hatch is provided to facilitate inspections and the maintenance of spargers and other hardware inside the tank.

• The COL applicant is to provide the operational procedures and maintenance program related to leak detection and contamination control (COL 6.8(1)). Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

• The COL applicant is to provide the preparation of cleanliness, housekeeping, and foreign materials exclusion program (COL 6.8(2)). This program addresses other debris sources, such as latent debris, inside containment. This program minimizes foreign materials in the containment.

• The COL applicant is to maintain the complete documentation of system design, construction, design modifications, field changes, and operations (COL 6.8(3)). Documentation requirements are included as a COL Information item.

12.4-66 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (17 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 6 Site Radiological

• The IRWST is designed to provide safety features for uninterrupted operation of the plant. 6.8.2 Environmental Monitoring Through monitoring, in-service inspection, and lessons learned from industry experiences, the integrity of the IRWST is well maintained, and contributes a very low level of contamination to the facility. Because of its location inside containment, the potential for environmental contamination of soil and groundwater from liquid leakages is minimized. Hence, this system is not required to be included in the site radiological environmental monitoring program.

12.4-67 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (18 of 75)

System: Chemical and Volume Control System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The system design, including the VCT, the EDT, and the letdown purification equipment inside 9.3.4 of Unintended the auxiliary building (AB); the RDT located inside the reactor containment building; and the Contamination holdup tank, BAST, and RMWT located outside the auxiliary building are designed with stainless steel and carbon steel materials that are compatible with the chemical and radiological environment. All components are of welded construction for life-cycle planning, thus minimizing leakage and unintended contamination of the facility and the environment.

• The CVCS is designed with provisions to collect drainage from system components, including pump seal leaks, valve leaks, and relief valve leakages, for purification and recycle. This design approach minimizes unintended contamination of the facility and the environment.

• The heat exchangers inside the containment are designed to minimize the potential for leakage by using shell-and-tube type designs that comply with ASME Section III.

• The piping tunnel connection from the yard tanks to the AB is designed to seismic Category I, constructed with low-porosity concrete, and to has above-ground building entrance and penetrations, thereby minimizing underground piping penetrations. This design approach minimizes the potential for the infiltration of water and the spread of contamination to the environment. Additionally, yard tank overflows are routed back to the AB sump for collection and processing to minimize unintended contamination.

12.4-68 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (19 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 (cont.)

• Drains from cleaning and maintenance activities are collected and routed for treatment in the DE system.

2 Adequate and Early Leak

• The CVCS is equipped with instrumentation to assist the detection of RCS leakage. Variations 9.3.4 Detection of VCT level, RDT level, EDT level, letdown flow rate, and charging flow rate indicate the potential for RCS leakage.

• The VCT, RDT, EDT, and CVCS processing equipment is designed with sufficient capacity to provide temporary holding of primary coolant for normal operation, including anticipated operational occurrences. The tanks are equipped with level indicating and control instruments to prevent overflow and provide reasonable assurance of a timely processing, thus minimizing interruption of normal processing operation, the spread of contamination, and waste generation.

• The sumps in the cubicles in which the CVCS tanks are located are designed with leak detection instruments to initiate alarms for operator actions in the event of leakage or overflow. The leak detection design has the capability to detect a small quantity of leakage to provide early warning to operators.

• The piping connecting the outside tanks to the SSCs that are inside the AB is routed through a pipe tunnel. The floor of the piping tunnel is sloped and epoxy coated, and has a sump with a level switch to detect piping leakage and infiltrated water. In the event that liquid is detected, the level switch sends a signal to the MCR for operator action. This design approach minimizes contamination of the environment.

12.4-69 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (20 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 3 Reduction of Cross-

• The SSCs are designed with life-cycle planning through the use of nuclear industry-proven 9.3.4 Contamination, materials compatible with the chemical, physical, and radiological environment, thus minimizing Decontamination, and waste generation.

Waste Generation

• The primary coolant purity is maintained with chemical injection for optimal reactor operation.

Boric acid is added to compensate for reactivity changes, fuel burnup, and xenon variations, and to provide shutdown margin. Lithium hydroxide is added for pH control, and hydrogen is added in order to minimize the occurrence of radiolysis. The addition of chemicals helps to minimize the generation of contaminated waste.

• Boric acid in the letdown flow is recovered for reuse to the maximum extent possible. In the event that the boric acid concentrate contains an abnormal quantity of radioactivity, the concentrate is sent to the liquid waste management system (LWMS) for neutralization, treatment, and release. The boric acid concentrator operates automatically to the desired boron concentration and sends the concentrate to the BAST for reuse. This design approach minimizes waste generation.

• The holdup tank, BAST, and the RMWT are located outside in a tank house designed to prevent the infiltration of rainwater and the spread of contamination. The tank house is designed with a sloped floor that is coated with epoxy to facilitate draining and cleaning, and is equipped with a sump that has level switch instrumentation to detect leakage and overflows. In the event that leakage is detected, the level switch sends a signal to the MCR for operator actions. This design approach minimizes the spread of contamination and waste generation.

• The process piping containing contaminated fluids is properly sized to facilitate flow with sufficient velocities to prevent the settling of solids. The piping is designed to reduce fluid traps, thus reducing the decontamination needs and waste generation. Decontamination fluid is collected and routed to the LWMS for processing and release.

• Utility connections are designed with a minimum of two barriers to prevent the spread of contamination to clean systems.

12.4-70 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (21 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 4 Decommissioning

• The SSCs are designed with decontamination capabilities. Design features, such as integrated 9.3.4 Planning component packages, utilized welding techniques used, and surface finishes are intended to minimize the need for decontamination and the resulting waste generation.

• The SSCs are designed for the full service life of the plant and are fabricated as individual assemblies for easy removal to the maximum extent practicable.

• The CVCS is designed with minimal embedded and buried piping. Piping between buildings is equipped with piping sleeves so that leakage is directed back to the building for collection, thus preventing unintended contamination.

5 Operations and

• The CVCS is designed for automated operation with manual initiation. Boron injection is 9.3.4 Documentation controlled by the makeup subsystem to maintain the desired boron concentration. Adequate instrumentation, including level, flow, and pressure elements, as well as process sampling, is provided to monitor and control the CVCS operations to prevent undue interruption, thus minimizing the spread of contamination and waste generation.

• Leak detection instruments are provided to detect individual tank leakage. Adequate space is provided around the equipment to enable prompt assessment and responses when required.

• Operational procedures and maintenance programs as related to leak detection and contamination control are to be prepared by the COL applicant (COL 9.3(1)). Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

• Complete documentation of system design, construction, design modifications, field changes, and operations is to be maintained by the COL applicant (COL 9.3(2)). Documentation requirements are included as a COL information item.

6 Site Radiological

• The CVCS is part of the overall plant and is included in the site radiological environmental 9.3.4 Environmental Monitoring monitoring program for monitoring of facility and environmental contamination. The program includes sampling and analysis of boric acid and contaminated fluid leakage from the BAST, RMWT, holdup tank, and piping tunnel, meteorological conditions, hydrogeological parameters, and potential migration pathways of radioactive contaminants. The program is included as a COL information item (COL 9.3(3)).

12.4-71 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (22 of 75)

System: Steam Generator Blowdown System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The SGBS components are located in elevated cubicles inside the auxiliary building. The 10.4.8 of Unintended cubicle floors are sloped, coated with epoxy, and provided with drains that are routed to the local Contamination drain hubs. This design approach prevents the spread of contamination through the facility and to the environment.

• The SGBS is designed with sufficient capacity for different modes of operation, including CBD, ABD, HCBD, and emergency BD. The system piping is adequately sized to prevent blockage and is sloped to facilitate drainage and prevent crud buildup.

• The flash tank, heat exchangers, filters, demineralizers, and the wetted parts of the WLS recirculation pumps are fabricated from stainless steel material and use welded construction for life-cycle planning, thus minimizing leakage and unintended contamination of the facility and the environment.

• The facility areas that house the system components, including the equipment cubicles that contain radioactively contaminated or potentially contaminated fluid, are designed with sloped floors with epoxy coating to facilitate the draining of fluid into drain pipes that direct liquid into a local sump. The SGBS component cubicles have epoxy-coated walls to facilitate cleaning.

The facility layout facilitates the operators' prompt assessment and fast response when needed.

2 Adequate and Early Leak

• The SGBS is designed with automated operation with manual initiation for the different 10.4.8 Detection modes of operation. Adequate instrumentation, including level, flow rate, temperature, and pressure elements, and a process radiation monitor, is provided to monitor the system operation to prevent undue interruption.

3 Reduction of Cross-

• The SSCs are designed with life-cycle planning through the use of nuclear industry-proven 10.4.8 Contamination, materials compatible with the chemical, physical, and radiological environment, thus minimizing Decontamination, and waste generation.

Waste Generation

• The SGBS components are provided with demineralized water for decontamination. Nitrogen and other utilities are provided to facilitate operations. The utility connections are designed with a minimum of two barriers to prevent contamination of clean systems.

12.4-72 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (23 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 3 (cont.)

• Process sampling connections are provided to determine the levels of contamination, treatment requirements, and confirmation of the continual radiation monitoring output. Continuous process radiation monitoring is provided on the outlet line of the treated blowdown water.

The detection of high radiation levels initiates automatic valve closure for isolation and operator actions, minimizing cross-contamination.

4 Decommissioning

• The SSCs are designed for the full service life and are fabricated as individual assemblies 10.4.8 Planning for easy removal to the maximum extent possible.

• The SSCs are designed to facilitate decontamination. Design features, such as the welding techniques used and surface finishes, minimize the need for decontamination and the resultant waste generation.

• The SGBS is designed without any embedded or buried piping. Piping between buildings is equipped with piping sleeves or tunnel, as applicable, with leak detection features, thus preventing contamination of the environment.

5 Operations and

• The removal and packaging of spent filter elements and spent resin are designed for remote 10.4.8 Documentation manual operation. Adequate space is provided around the equipment to enable prompt assessment and responses.

• The COL applicant is to establish operational procedures and maintenance programs as related to leak detection and contamination control (COL 10.4(1)). Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

• The COL applicant is to maintain complete documentation of the system design, construction, design modifications, field changes, and operations (COL 10.4(2)). Documentation requirements are included as a COL information item.

6 Site Radiological

• The SGBS is part of the overall plant and is included in the site radiological 10.4.8 Environmental Monitoring environmental monitoring program for monitoring the potential for environmental contamination. The program includes sampling and analysis of waste samples, meteorological conditions, hydrogeological parameters, and potential migration pathways of the radioactive contaminants. The COL applicant is to prepare the Site Radiological Environmental Monitoring Program (COL 10.4(8)).

12.4-73 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (24 of 75)

System: Component Cooling Water System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The system components, including the CCWS surge tanks, pumps, and associated piping, are 9.2.2 of Unintended fabricated from carbon steel material and are of welded construction for life-cycle planning.

Contamination The CCWS is injected with a corrosion inhibitor, and the surge tanks are covered with nitrogen gas to minimize the oxidation of steel and formation of corrosion. The CCWS heat exchangers are made of titanium, and pump impellers are made of stainless steel. The system design includes periodic sampling and analysis to maintain water quality and provisions for in-service testing and inspection to maintain system integrity. This design approach minimizes leakage and unintended contamination of the facility and the environment.

• The CCWS surge tanks, heat exchangers, and pumps are designed with multiple divisions and sufficient capacity to accommodate different modes of operation, including normal and anticipated operational occurrences. The tanks are designed to meet ASME III, Division I, and are equipped with dual-level instruments to facilitate control of the content liquid level, thus minimizing the spread of contamination and waste generation.

• The facility area that houses the system components, including the equipment cubicles that contain radioactively contaminated or potentially contaminated fluid, are designed with sloped floors with epoxy coating to facilitate the draining of fluid into drain pipes that direct liquid into a local sump. The facility layout facilitates the operators prompt assessment and fast response when needed.

• Sumps that contain contaminated or potentially contaminated fluid are equipped with stainless steel liners and level switches to initiate pumping when sump levels reach a predetermined setpoint. The sumps are designed to facilitate periodic maintenance and inspection of the epoxy coating.

12.4-74 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (25 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 2 Adequate and Early Leak

• CCWS surge tanks are designed with dual-level instruments to provide reasonable assurance of 9.2.2 Detection safe operation and to provide alarms to the operating personnel in the event of overflow or leakage.

• Two radiation monitors, one for each division, are provided to continuously monitor contamination levels at the discharge of the CCWS pumps and indicate radiation activity in the MCR. In the event that radiation is detected above the pre-determined limit, an alarm is initiated by one of the monitors for operator action. Because the CCWS is segregated into two independent and parallel divisions, each division can be isolated for inspection, mitigation, and maintenance. This design approach provides early leak detection and minimizes the spread of contamination to other components, the facility, and the environment.

3 Reduction of Cross-

• SSCs are designed with life-cycle planning through the use of nuclear industry- proven 9.2.2 Contamination, materials that are compatible with the chemical, physical, and radiological environment, thus Decontamination, and minimizing waste generation.

Waste Generation

• CCWS surge tanks are located at a high elevation to maintain liquid pressure in order to prevent infiltration and cross-contamination of the CCWS. This location also helps to prevent unintended contamination of the environment because any leaks from the tanks are drained to the local sumps for collection and forwarded to the LWMS for treatment and disposal.

• The cubicles where the CCWS SSCs are housed are designed with sloped floors, epoxy coating to provide drainage and cleanable surfaces, and local sumps to collect leakages and overflows.

Cubicle curbs are also provided to reduce cross- contamination and the spread of contamination to other areas. CCWS heat exchangers are located in a separate structure that is close to the essential service water building in order to minimize radiation exposure to the essential service water piping.

• Plate-type heat exchangers are used to transfer heat from the component cooling water side to the essential service water side. The heat exchangers are designed with titanium plates to minimize the potential for pinhole leaks between the component cooling water and essential service water. The gaskets between the plates are also designed to direct leakages to the outside of the heat exchangers. Leakages are then collected in the floor drain sumps and routed for treatment and release. This design approach minimizes contamination of clean systems as well as to the facility and the environment.

12.4-75 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (26 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 3 (cont.)

• Sampling points are provided to facilitate sampling and analyses to provide reasonable assurance that the water quality is maintained and also to determine the need for the addition of corrosion inhibitor. This design approach minimizes waste generation.

• Utility connections (e.g., demineralized water, nitrogen) are designed with a minimum of two barriers to prevent the contamination of clean systems.

4 Decommissioning

• SSCs are designed for the full service life and are fabricated as individual assemblies for easy 9.2.2 Planning removal to the maximum extent possible.

• The CCWS is designed with minimum embedded or buried piping. Piping between buildings is equipped with piping sleeves or tunnels, as applicable, with leak detection features. Piping to the component cooling water heat exchanger structures is routed through a seismic Category I reinforced concrete pipe tunnels (one per division) under the yard. The tunnel is coated with epoxy and is designed with a collection sump with a level switch to initiate an alarm signal in the MCR for operator actions in the event liquid is detected. This design approach thus minimizes unintended contamination of the facility and the environment.

5 Operations and

• The CCWS is designed for automated operation with manual initiation for the different 9.2.2 Documentation modes of operation. CCW surge tanks are designed with dual-level instruments to provide reasonable assurance of safe operation. A high-level signal, indicating in-leakage, or a low-level signal, indicating out-leakage, is transmitted to the MCR for operator action. A low-low-level signal isolates the nonessential headers and the RCP headers from the remaining portion of the system, thus minimizing the spread of contamination and waste generation.

• Adequate space is provided around the equipment to enable prompt assessment and responses as needed.

• 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(10)). Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

• The COL applicant is to maintain complete documentation of system design, construction, design modifications, field changes, and operations in the CCWS (COL 9.2(11)).

Documentation requirements are included as a COL information item.

12.4-76 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (27 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 6 Site Radiological

• The CCWS is considered to have a potential low level of contamination through leakage from 9.2.2 Environmental Monitoring contaminated systems through heat exchangers. Through monitoring, inservice inspection, and lessons learned from industry experience, the integrity of the CCWS is maintained, resulting in a minimal level of contamination. Leakage from the system to the facility and the environment is captured by the drainage collection provisions. Any residual contamination of the hydrogeology is not likely to be distinguishable from other contamination sources. Hence, contamination characteristics of the CCWS are not monitored in the site-wide program.

However, the COL applicant is to include a site-wide radiological environmental monitoring program to monitor environmental contamination in the CCWS (COL 9.2(12)).

12.4-77 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (28 of 75)

System: Essential Service Water System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The ESW facility is designed not to be in close contact with radioactively contaminated 9.2.1 of Unintended components and is located away from contamination areas to prevent unintended contamination.

Contamination

• All openings at the pump house floor are sealed to prevent water entry and preclude the flooding of the ESW pumps and other safety-related equipment within the structure.

2 Adequate and Early Leak

• Radiation monitors are provided to detect contamination after the exchanging heat with CCW 9.2.1 Detection heat exchangers. This design approach provides early detection of contamination.

• The ESWS is designed to be readily accessible for inspection and maintenance.

3 Reduction of Cross-

• The heat exchangers are constructed with titanium material to minimize pinhole leaks. 9.2.1 Contamination,

• The heat exchanger seals are designed to leak toward the outside of the heat exchangers where Decontamination, and leakage is collected in the building sump.

Waste Generation

• A sump is provided for collection of any leakage. The sump is designed with steel liners and is equipped with level instruments to initiate an alarm signal for operator actions.

4 Decommissioning

• The ESWS is designed for the full service life and is fabricated as individual assemblies for easy 9.2.1 Planning removal.

• The ESWS is designed with minimal embedded or buried piping. Piping between buildings is designed to be routed in seismic Category I, reinforced concrete pipe tunnels (one per division) under the yard.

5 Operations and

• The ESWS is designed for automated operations with manual initiation for the different modes of 9.2.1 Documentation operation in conjunction with the CCWS.

• Adequate ingress and egress spaces are provided for prompt assessments and appropriate responses when and where they are needed.

12.4-78 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (29 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 6 Site Radiological

• The ESWS is designed to minimize the potential for contamination through leakage in the heat 9.2.1 Environmental Monitoring exchangers. Through monitoring, in-service inspection, and lessons learned from industry experiences, the integrity of the ESW/CCW heat exchangers is expected to be well maintained, resulting in no contamination or to a very low level of contamination of the system. Leakage from the system to the facility and the environment is captured by the design. Any residual contamination of the hydrogeology is not likely to be distinguishable from other contamination sources. Hence, ESWS has low risk and low radiological consequence, and this design is in compliance with RG 4.21.

12.4-79 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (30 of 75)

System: Spent Fuel Pool Cooling and Cleanup System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The SFPCCS components are located in elevated cubicles inside the auxiliary building. 9.1.3 of Unintended The cubicle floors are sloped and coated with epoxy and are provided with drains that are routed Contamination to the local drain hubs. This design approach prevents the spread of contamination within the facility and to the environment.

• The spent fuel pool, refueling pool, refueling canal, and cask loading pit are installed with stainless steel liner plates and welded seam drain channels. Other components, including heat exchangers, filters, demineralizers, and pumps, are fabricated from stainless steel material and utilize welded construction for life- cycle planning, thus minimizing leakage and unintended contamination of the facility and the environment.

2 Adequate and Early Leak

• The SFPCCS is designed to include sight glasses for visual inspections. Early leak detection 9.1.3 Detection can be achieved through the use of sight glasses and operating procedures.

• Adequate instrumentation, including level, flow rate, temperature, and pressure elements, is provided to monitor the system operation to prevent undue interruption.

3 Reduction of Cross-

• The SSCs are designed with life-cycle planning through the use of nuclear industry-proven 9.1.3 Contamination, materials that are compatible with the chemical, physical, and radiological environment, thus Decontamination, and minimizing waste generation.

Waste Generation

• Plate-type heat exchangers are used for thermal transfer from the spent fuel pool cooling side to the component cooling water side. The heat exchangers are designed with stainless steel plates to minimize the potential for pinhole leaks between the potentially contaminated system and component cooling water. The heat exchangers are designed for the component cooling water to operate at a higher pressure than the process fluid side, which prevents the spread of contamination to the clean system side through leakage. The gaskets between the plates are also designed to direct leakage to the outside of the heat exchangers. The leakage is then collected in the local floor drain sump. This design approach minimizes the spread of contamination to the facility and the environment.

• The utility connections are designed with a minimum of two barriers to prevent the contamination of clean systems.

12.4-80 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (31 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 4 Decommissioning

• The SSCs are designed for the full service life and are fabricated as individual assemblies 9.1.3 Planning for easy removal, with the exception of the liner plates.

• The SSCs are designed to facilitate decontamination. Design features, such as the welding techniques that are used and surface finishes, are included to minimize the need for decontamination and the resultant waste generation.

• The SFPCCS is designed with minimum embedded or buried piping. Piping between buildings is equipped with piping sleeves or tunnel, as applicable, with leak detection features, thus preventing unintended contamination to the environment.

5 Operations and

• The removal and packaging of spent filter elements and spent resin is designed for remote manual 9.1.3 Documentation operation. Adequate space is provided around the equipment to enable prompt assessment and responses when required.

• 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(1)). Procedures and maintenance programs are to be completed before fuel is loaded.

• The COL applicant is to maintain complete documentation of system design and system design modifications (COL 9.1(7)). Documentation requirements are included as a COL information item.

12.4-81 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (32 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 6 Site Radiological

• The SFPCCS is designed to manage radioactive contamination through the storage of spent fuel. 9.1.3 Environmental Monitoring The integrity of the SFPCCS is maintained through monitoring, in- service inspection, and the implementation of lessons learned from industry experience. Maintaining the SFPCCS results in a low level of contamination in the facility. Because the SFPCCS is located at higher plant elevations, the potential for environmental contamination of soil and groundwater from pool liquid leakages is minimized. However, because the pool is open, contamination from the evaporation of water from the SFPCCS and other systems is included in the site radiological environmental monitoring program. The program is included as a COL information item.

12.4-82 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (33 of 75)

System: Gaseous Waste Management System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The system contains sufficient charcoal material to hold the noble gas nuclides for a period of 11.3.2 of Unintended decay to reduce the release of radioactivity, thus minimizing contamination of the facility and the Contamination environment.

• The system design, including the waste gas dryers and guard beds, is configured in two parallel trains, one operating and one standing by, each with sufficient capacity to remove the moisture to protect the charcoal beds.

• A HEPA filter is provided downstream of the charcoal beds to prevent the spread of contaminated charcoal fines.

• Cubicles in which contaminated materials are stored and processed are epoxy-coated to facilitate cleaning. The GRS header drain tank is equipped with level instrumentation to detect fluid accumulation and drain the fluid to the radioactive drain system.

• The system is designed with above-ground piping to the extent practicable. Buried and embedded piping is minimized. Piping is sloped to facilitate drainage of condensate to the header drain tank.

• The system uses valves with leak-tight characteristics, such as the bellows or metal diaphragm types, to minimize leakage.

• The system uses welded construction to the maximum practicable extent to minimize leakage.

2 Adequate and Early Leak

• The system is designed with gas analyzers and a radiation monitor to provide reasonable 11.3.2 Detection assurance of the integrity of the SSCs, including piping, and to provide alarms to warn operators of the potential for explosive gas concentrations.

• The system is designed with adequate space around all components to enable prompt evaluation and response to leakage detection.

3 Reduction of Cross-

• The SSCs are designed with life-cycle planning through the use of nuclear-industry-proven 11.3.2 Contamination, equipment and materials that are compatible with the chemical, physical, and radiological Decontamination, and environment, thus minimizing cross-contamination and waste generation.

Waste Generation

• The process piping containing contaminated fluid is sloped to facilitate flow and reduce fluid traps, thus reducing decontamination and waste generation. Decontamination fluid is collected and routed to the LWMS for processing and release.

12.4-83 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (34 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 3 (cont.)

• Utility connections are designed with a minimum of two barriers to prevent contamination of nonradioactive systems from radioactive systems.

4 Decommissioning

• The SSCs are designed for the full service life and are fabricated, to the maximum extent 11.3.2 Planning practicable, as individual assemblies for easy removal.

• The SSCs are designed with decontamination capabilities using low-pressure nitrogen. Design features such as welding techniques and surface finishes are included to minimize the need for decontamination and minimize waste generation.

• The GRS is designed with minimal embedded or buried piping. The drain gas header between buildings is equipped with piping sleeves with leakage directed back to the compound building for collection, thus preventing the spread of contamination.

5 Operations and

• The GRS is designed for remote and automated operations. The system is equipped with 11.3.2 Documentation instruments to actuate the drain valve from the header drain tank.

• The COL applicant is to prepare the operational procedures and maintenance programs related to leak detection and contamination control (COL 11.3(4)). Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

• The COL applicant is to maintain the complete documentation of system design, construction, design modifications, field changes, and operations (COL 11.3(5)). Documentation requirements are included as a COL information item.

6 Site Radiological

• The GRS is part of the plant and is included in the site process control program and the site 11.3.2 Environmental Monitoring radiological environmental monitoring program for monitoring of facility and environmental contamination. The site radiological environmental monitoring program includes sampling and analysis of effluent to be released, meteorological conditions, hydrogeological parameters, and potential migration pathways of radioactive contaminants. The COL applicant is to prepare the site process control program and the site radiological environmental monitoring program (COL 11.3(6)).

12.4-84 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (35 of 75)

System: Liquid Waste Management System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The system components, including the collection tanks and the monitor tanks, are fabricated of 11.2.2 of Unintended stainless steel material and are of welded construction for life-cycle planning, thus minimizing Contamination leakage and unintended contamination of the facility and the environment.

• The LWMS tanks are designed with sufficient capacity to provide temporary storage of the liquid waste generated from normal operation including anticipated operational occurrences. The tanks are equipped with cross-connected inlet headers for the control of overflow and to provide reasonable assurance of timely processing. The design minimizes the interruption of normal processing operation, the spread of contamination, and waste generation.

• The LWMS tanks are designed with mixing eductors to minimize settling of suspended solids.

The tanks have polished internal surfaces to minimize crud traps.

• The LWMS is designed with minimum embedded or buried piping. Piping between buildings is designed to be equipped with piping sleeves with leakage directed back to the building for collection, thus preventing the spread of contamination.

2 Adequate and Early Leak

• All LWMS tanks are designed with level instruments to provide reasonable assurance of safe 11.2.2 Detection operation of the SSCs. The instruments provide alarms to the main control room and the radwaste control room for operator action in the event of high liquid levels.

• The cubicles in which the LWMS tanks are located are designed to include leak detection instrumentation to initiate alarms for operator actions in the event of leakage. The leak detection design has the capability to detect a small quantity of leakage for early detection.

12.4-85 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (36 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 3 Reduction of Cross-

• The SSCs are designed with life-cycle planning through the use of nuclear industry-proven 11.2.2 Contamination, materials compatible with the chemical, physical, and radiological environment, thus minimizing Decontamination, and waste generation.

Waste Generation

• The floor drains, equipment drains, chemical drains, and the detergent wastes are collected in segregated tanks in separate cubicles. Because the LWMS is designed to operate in batches, treatment for these collected wastes is determined through sampling and analyses. This design approach minimizes cross-contamination and waste generation.

• The process piping containing contaminated solids is sized to facilitate flow and to provide for velocities that are sufficient to prevent the settling of solids. The piping is designed to reduce crud traps, thus reducing decontamination and waste generation. Decontamination fluid is collected and processed.

• Utility connections are designed with a minimum of two barriers to prevent contamination of nonradioactive systems from potentially radioactive systems.

4 Decommissioning

• The SSCs are designed with decontamination capabilities. Design features such as spargers, 11.2.2 Planning welding techniques, and surface finishes are included to minimize the need for decontamination and minimize waste generation.

• The SSCs are designed for the full service life and are fabricated, to the maximum extent practicable, as individual assemblies for easy removal.

5 Operations and

• The COL applicant is to maintain complete documentation of the system design, construction, 11.2.2 Documentation design modification, field changes, and operations (COL 11.2(10)).

6 Site Radiological

• The LWMS is included in the site process control program and the site radiological 11.2.2 Environmental Monitoring environmental monitoring program for monitoring facility and environmental contamination.

The COL applicant is to prepare the site process control program and the site radiological environmental monitoring program (COL 11.2(11)). The site radiological environmental monitoring program includes sampling and analysis of effluent to be released, meteorological conditions, hydrogeological parameters, and potential migration pathways of radioactive contaminants.

12.4-86 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (37 of 75)

System: Solid Waste Management System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The system components, including the low-activity spent resin tank and the spent resin long-term 11.4.2 of Unintended storage tank, are designed with stainless steel material and welded construction for life-cycle Contamination planning. The tanks are designed to have sufficient capacity to contain the spent resins for a decay period to reduce radioactivity.

• The concentrate treatment subsystem and the spent resin dewatering subsystem are designed as skid packages with self-containing drip pans to contain leakage. The drains connected to the drip pan are routed to a floor drain sump for collection and then pumped to the LWMS for treatment and release.

• The temporary waste storage area is designed with epoxy-coated floors and a drainage system to direct drainage to a floor drain sump for collection and subsequent pumping to the LWMS for treatment and release.

• Cubicles in which contaminated materials are handled and stored have floors that are sloped and epoxy-coated to facilitate cleaning and facilitate drainage to early leak detection piping. Sump tanks are equipped with level switches to detect liquid accumulation, and pumps are provided to transfer the fluid for proper treatment.

• The SWMS is designed with above-ground piping to the extent practicable. Buried and embedded piping is minimized. In lieu of embedded piping, a drainage pipe is provided in a covered concrete trench to collect any component overflow or leakage and route it to the building sump. In the event that buried or embedded piping cannot be avoided, double-walled piping and leak detection instruments are considered and evaluated based on the risks and the radiological consequences associated with the contamination of the facility and the environment.

2 Adequate and Early Leak

• The low-activity spent resin tank and the spent resin long-term storage tank are designed with 11.4.2 Detection level instrumentation to provide reasonable assurance of the integrity of the SSCs including the associated piping and to provide alarms to warn the operators of leakages.

• The temporary waste storage area is designed with remote cameras for waste handling operation as well as for periodic surveillance.

• Other components are designed for batch operation and are provided with adequate space to enable prompt assessment and response when required.

12.4-87 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (38 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 3 Reduction of Cross-

• The SSCs are designed with life-cycle planning through the use of nuclear industry-proven 11.4.2 Contamination, materials compatible with the chemical, physical, and radiological environment, thus minimizing Decontamination, and cross-contamination and waste generation.

Waste Generation

• The process piping containing contaminated slurry is sized to facilitate flow and provide velocities that are sufficient to prevent solids settling in accordance with ANSI/ANS-55.6. The piping is designed to reduce fluid traps, thus reducing decontamination and waste generation per 10 CFR 20.1406. Pipe flushing fluid is collected and routed to the LWMS for processing and release, per ANSI/ANS-55.1.

• Utility connections are designed with a minimum of two barriers to prevent contamination of nonradioactive systems from potentially radioactive systems.

4 Decommissioning

• The SSCs are designed for the full service life and are fabricated, to the maximum extent 11.4.2 Planning practicable, as individual assemblies for easy removal.

• The SSCs are designed with decontamination capabilities. Design features such as welding techniques and surface finishes are included to minimize the need for decontamination and minimize waste generation.

5 Operations and

• The spent resin removal and packaging are designed for remote and automated operations. The 11.4.2 Documentation system is equipped with instruments to provide alarms for operator actions in the event leakage is detected.

• The operational procedures and maintenance programs related to leak detection and contamination control are to be prepared by the COL applicant (COL 11.4(10)). Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

• A RG 4.21 program, Life Cycle Planning of Minimization of Contamination, is to be developed by the COL applicant to identify plant-wide components that contain radioactive materials, buried piping, embedded piping, leak detection methods and capabilities, and the methods utilized for the prevention of unnecessary contamination of clean components, facility areas, and the environment following the guidance of NEI 08-08A (COL 11.4(11)). The leak identification program for the SWMS is to be integrated into this plant-wide program to facilitate the timely identification of leaks, prompt assessment, and appropriate responses to isolate and mitigate leakage.

12.4-88 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (39 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 5 (cont.)

• Complete documentation of the system design, construction, design modifications, field changes, and operations is to be maintained by the COL applicant. Documentation requirements are included as a COL information item (COL 11.4(12)).

6 Site Radiological

• The SWMS is part of the plant and is included in the site process control program and the site 11.4.2 Environmental Monitoring radiological environmental monitoring program for monitoring of facility and environmental contamination. The site radiological environmental monitoring program includes sampling and analysis of effluent to be released, meteorological conditions, hydrogeological parameters, and potential migration pathways of radioactive contaminants. Both programs are included as COL information items (COL 11.4(5)).

12.4-89 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (40 of 75)

System: Radioactive Drain System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The system components, including all of the EFDS sumps that handle radioactive or potentially 9.3.3 of Unintended radioactive liquid, are designed with stainless steel liners encased in concrete cavities. The Contamination concrete cavities are epoxy-coated to provide smooth surfaces for cleaning and decontamination for life-cycle planning. The EFDS sump liners are designed to be sealed at the rim to prevent infiltration of contaminated drainage. The sump liners are also designed to be removable to facilitate cleaning, inspection, and maintenance activities on a predetermined frequency based on the service life of the epoxy coating. This design approach minimizes leakage and unintended contamination of the facility and the environment.

• The EFDS sumps are equipped with duplex pumps and level instruments to facilitate automated pump starting and stopping operation. Normally one pump in the sump is used. In the event of excessive collected drainage, the second pump in the sump is automatically started on the high-high-level signal. This design approach minimizes sump overflow and thus minimizes the spread of contamination and waste generation.

2 Adequate and Early Leak

• In areas that are not normally accessible during power operation, such as the ICI cavity sump, the 9.3.3 Detection RCB drain sump, and the AB floor drain sumps, continual level indication is provided in the MCR and the RSR. In the event the sump liquid level reaches a preset limit, the level instrument initiates a signal to alarm in the MCR and the RSR for operator actions. This design approach, supplemented with operational procedures, provides adequate leak detection capability and minimizes the potential for the spread of contamination.

3 Reduction of Cross-

• The sumps that handle contaminated or potentially contaminated fluids are designed to minimize 9.3.3 Contamination, the spread of contamination through the use of liners, epoxy coating, and seals. This design Decontamination, and approach is nuclear-industry proven and is compatible with the chemical, physical, and Waste Generation radiological environment, thus minimizing waste generation.

• The EFDS sumps are strategically located to collect floor and equipment drainage. Drains are segregated for different handling and processing requirements to minimize cross-contamination.

The pumps are adequately sized for maximum input to prevent overflow; one pump is normally used, but when the sump liquid content level continues to rise, the second pump is initiated at the high-high-level. This design approach minimizes the spread of contamination.

12.4-90 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (41 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 3 (cont.)

• The sumps are housed in individual cubicles. Cubicle curbs and/or doors are provided to reduce the spread of contamination to other areas and additional needs for decontamination.

• Sampling points are provided to facilitate sampling and analyses to properly route drainage for treatment and disposal. This design approach minimizes waste generation.

• Utility connections such as demineralized water and instrument air are designed with a minimum of two barriers to prevent cross-contamination.

4 Decommissioning

• The sump liners are designed for the full service life and are fabricated as individual assemblies 9.3.3 Planning for easy removal. The drain pipe flange connections to the sumps are also designed to permit disconnect to facilitate removal of the liners for inspection, cleaning, and maintenance of the epoxy coating of the concrete cavity. This design approach minimizes contamination and facilitates decommissioning.

• The EFDS is designed with minimum embedded or buried piping through the use of pump transfer. Drain pipes are routed through pipe chases as much as practicable. Piping between buildings is equipped with piping sleeves with leakage directed back to the originating facility for collection. This design approach minimizes the spread of contamination and facilitates decommissioning.

5 Operations and

• The EFDS is designed for automated operation. The sumps are designed with dual-level 9.3.3 Documentation instruments to provide reasonable assurance of a safe operation. A high level signal initiates liquid transfer with one pump. If the liquid level continues to rise, a high-high signal initiates additional liquid transfer with the secondary pump. Level signals are transmitted to the MCR for monitoring and operator response.

• Adequate space is provided in the vicinity of the sumps to enable prompt assessment and responses when required.

• The COL applicant is to provide operational procedures and maintenance programs as related to leak detection and contamination control (COL 9.3 (1)). Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

• The COL applicant is to maintain complete documentation of the system design, construction, design modifications, field changes, and operations (COL 9.3 (2)). Documentation requirements are included as a COL information item.

12.4-91 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (42 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 6 Site Radiological

• The EFDS is designed to handle radioactive fluids, and the sumps are located at lower elevations 9.3.3 Environmental Monitoring to facilitate drainage collection. The EFDS is part of the overall plant and is included in the site radiological environmental monitoring program for monitoring of facility and environmental contamination. The site radiological environmental monitoring program includes sampling and analysis of effluent to be released, meteorological conditions, hydrogeological parameters, and potential migration pathways of radioactive contaminants. The COL applicant is to prepare the Site Radiological Environmental Monitoring Program (COL 9.3 (3)).

12.4-92 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (43 of 75)

System: Primary Sampling System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The NPSS and SSS components for normal operation, including anticipated operational 9.3.2 of Unintended occurrences, are located in elevated cubicles inside the compound building. The normal primary Contamination and secondary sampling cooler racks and the associated sinks are equipped with drain lines that are connected to the local drain piping. A hood is provided to remove radioactive gases from the collected samples during analysis to minimize the spread of contamination. In addition, the cubicle floors are sloped, coated with epoxy, and provided with drains that are routed to the local drain hubs. This design approach prevents unintended contamination of the facility and the environment.

• The post-accident sampling cooler rack and sink are located at the basemat level in the auxiliary building. The post-accident sampling rack and the sink are equipped with drain lines routed to the local drain piping. A hood is provided to remove radioactive gases from the collected samples during analysis to minimize the spread of contamination. In addition, the cubicle floors are sloped, coated with epoxy, and provided with drains that are routed to the local drain hubs.

This design approach prevents unintended contamination of the facility and the environment.

• The NPSS, PASS and SSS are designed to be segregated with individual coolers and sampling points. Sampling piping sizes and components are designed to provide desired flow and sampling conditions in order to obtain representative samples.

• The normal primary sample sink, normal primary sample cooler rack, secondary sample sink, secondary sample cooler rack, post- accident primary sample sink, and post-accident primary sample cooler rack are constructed of stainless steel material. The primary off-gas sample pump is fabricated from stainless steel material and is of welded construction for life-cycle planning, thus facilitating decontamination and minimizing the spread of contamination through leakage.

• The facility area that houses the system components, including the equipment cubicles that contain radioactively contaminated or potentially contaminated fluid, are designed with sloped floors with epoxy coating to facilitate the draining of fluid into drain pipes that direct liquid into a local sump. The equipment cubicles for the NPSS, PASS and SSS have epoxy coated wall to facilitate cleaning. The facility layout facilitates the operators' prompt assessment and fast response when needed.

12.4-93 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (44 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 2 Adequate and Early Leak

• The NPSS, PASS and SSS are designed with automated operation with manual initiation for the 9.3.2 Detection two modes of operation. Adequate instrumentation is provided to control the sampling operations, thus minimizing waste generation.

3 Reduction of Cross-

• The SSCs are designed with life-cycle planning through the use of nuclear industry-proven 9.3.2 Contamination, materials compatible with the chemical, physical, and radiological environment, thus minimizing Decontamination, and cross-contamination and waste generation.

Waste Generation

• The NPSS, PASS and SSS components are provided with demineralized water and nitrogen gas for decontamination and purging. The utility connections are designed with a minimum of two barriers to prevent the contamination of clean systems.

4 Decommissioning

• The SSCs are designed for the full service life and are fabricated as individual assemblies for easy 9.3.2 Planning removal. The components are located in accessible areas for operation, maintenance, and decommissioning purposes.

• The SSCs are designed with decontamination capabilities. Other design features, such as the welding techniques that are used and surface finishes, are implemented in order to minimize the need for decontamination and the resultant waste generation.

• The NPSS, PASS and SSS are designed without any embedded or buried piping. Piping between buildings is equipped with piping sleeves or tunnel, as applicable, with leak detection features, thus preventing unintended contamination of the environment.

5 Operations and

• The NPSS, PASS and SSS are designed for automated and remote operations with manual 9.3.2 Documentation initiation for the collection of samples. Adequate space is provided around the components to enable prompt assessment and response when required.

• The NPSS, PASS and SSS are a packaged vendor design. The COL applicant is to prepare operational procedures and maintenance programs as related to leak detection and contamination control (COL 9.3(1)). Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

• The COL applicant is to maintain complete documentation of system design, construction, design modifications, field changes, and operations (COL 9.3(2)). Documentation requirements are included as a COL information item.

12.4-94 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (45 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 6 Site Radiological

• The NPSS, PASS and SSS are designed to be used only for the purpose of collecting and 9.3.2 Environmental Monitoring analyzing samples. The procedures for proper handling and treatment of sampled materials include small volumes in the sampling operation. The NPSS, PASS and SSS are not included in the site radiological environmental monitoring program.

12.4-95 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (46 of 75)

System: Main Steam System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The main steam lines are equipped with radiation monitors to detect radiological contamination 10.3.2 of Unintended and to provide alarms for operator notification. Following a steam generator tube rupture, the Contamination MSIVs for the affected steam generator are closed in order to isolate the affected steam generator.

This design minimizes radiological cross-contamination and unintended contamination of the facility.

• The main steam piping is sloped in the direction of steam flow to avoid water entrenchment and the collection of condensate drainage.

• To minimize the possibility of water impingement into the main turbine, drain traps are provided at low points in the main steam piping where water may collect. Condensate from these drain traps is continuously removed with direct piping to the main condenser during normal plant power operation. This design approach prevents the spread of contamination within the facility.

2 Adequate and Early Leak

• Radiological monitoring of the MSS is provided continuously through N-16 gamma detection 10.3.2 Detection with a scintillation detector and microprocessor on each main steam line from the steam generator.

• Radiation monitors are provided on the condenser vacuum pump exhaust line and the steam generator blowdown line to monitor contamination levels associated with the condensate and steam generator blowdown systems.

• The radiation monitors are designed to provide indication of steam generator leakage and alert the operator of a steam generator tube leak condition during power operation, including the identification of the affected steam generator.

12.4-96 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (47 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 3 Reduction of Cross-

• The safety-related portion of the main steam piping is designed to comply with the ASME Code 10.3.2 Contamination, Section III. The non-safety-related portion of the main steam piping is designed to ASME B31.1 Decontamination, and for safe operation and the minimization of waste generation.

Waste Generation

• Main steam piping is designed to minimize the effects of erosion/corrosion, is adequately sized to limit velocities, and is routed with long-radius elbows to minimize potential erosion and waste generation.

• The SSCs are designed with life-cycle planning through the use of nuclear industry-proven materials compatible with the chemical, physical, and radiological environment, minimizing waste generation.

• Process sampling connections are provided downstream of the MSIVs for monitoring of steam chemistry.

4 Decommissioning

• The main steam piping is designed for the full service life and is fabricated as individual segments 10.3.2 Planning for easy assembly and removal.

• The main steam piping is designed with cleanout capabilities. Design features such as welding techniques used and surface finishes are intended to minimize the need for decontamination, and hence reduce waste generation.

• The MSS is designed without any embedded or buried piping, preventing contamination to the environment.

5 Operations and

• The MSS piping and components are located in the reactor containment building, auxiliary 10.3.2 Documentation building, and turbine generator building. Adequate space is provided around the equipment to enable prompt assessment and responses.

• The COL applicant is to establish operational procedures and maintenance programs as related to leak detection and contamination control (COL 10.3(2)). Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

6 Site Radiological

• The MSS is part of the overall plant but does not have direct release points or contamination 10.3.2 Environmental Monitoring migration pathways to the environment during normal operation. Therefore, the MSS is not specifically required to be included in the radiological environmental monitoring program.

12.4-97 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (48 of 75)

System: Condensate System and Feedwater System (includes Feedwater Pump Turbine System, Extraction Steam System and Heater Drain System)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The components for the condensate and feedwater system and the associated subsystems are 10.4.7 of Unintended located inside the turbine generator building (TGB). The floors are sloped, coated, and provided Contamination with drains that are routed to the TGB sumps. The drains from the TGB sumps are monitored for radiological contamination. Drainage that is detected to be contaminated is routed to the LWMS for treatment and release. The balance is routed to the wastewater treatment (WWT) facility for processing and release with other non-radiological wastewater. This design approach prevents the spread of contamination through the facility and to the environment.

• The condensate and feedwater systems and the associated subsystems are designed with sufficient capacities to accommodate different modes of operation. The piping is adequately sized to prevent blockage; primary process piping is sloped to facilitate drainage and prevent fluid accumulation and crud buildup.

• Process sampling connections are provided under each tube bundle in condenser hotwell to detect condenser tube leaks and determine location of condenser tube leaks.

• The feedwater heaters are designed to Heat Exchange Institute (HEI) and Tubular Exchanger Manufacturers Association (TEMA) Standards and heater tubes are manufactured using high-grade austenitic stainless steel for minimization of unintended leakage. The feedwater heaters are constructed of carbon steel and welded for life-cycle planning, thus minimizing leakage and unintended contamination of the facility and the environment.

• The heat exchangers are designed so that tube-side pressure is higher than shell-side pressure to protect against leakage from the potentially contaminated steam to the clean circulating water system.

• The feedwater pump turbine system is designed to exhaust the steam into the condenser, and to direct the casing leakage from the turbine to the condenser pit sump.

• To minimize the possibility of water impingement into the main turbine, drain traps are provided at low points in the extraction steam piping where water may collect. Condensate from these drain traps is continuously removed with direct piping to the main condenser during normal plant power operation. Drain pots are provided with level instrumentation that opens the low-point drain valves upon a high signal. This design approach prevents the spread of contamination within the facility.

12.4-98 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (49 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 (cont.)

• Extraction steam piping material is ferritic alloy steel (Cr-Mo) and is designed to minimize the effects of erosion / corrosion.

2 Adequate and Early Leak

• The condensate and feedwater systems and the associated subsystems are designed with 10.4.7 Detection automated operation with manual initiation for the different modes of operation. Adequate instrumentation, including level, flow rate, temperature, and pressure elements, is provided to monitor and control the operations to prevent undue interruption. This design approach minimizes the spread of contamination and waste generation.

• Radiation monitors are provided on the condenser vacuum pump exhaust line and the steam generator blowdown line to monitor contamination levels associated with the condensate and steam generator blowdown systems.

• Leak detection trays are included at all tube-to-tubesheet interfaces. Provisions for early leak detection are provided for the tubesheet trays and in each hotwell section.

3 Reduction of Cross-

• The SSCs are designed with life-cycle planning through the use of nuclear industry-proven 10.4.7 Contamination, materials compatible with the chemical, physical, and radiological environment, minimizing the Decontamination, and spread of contamination and waste generation.

Waste Generation

• Normal and emergency drains are routed to the condenser and are forwarded to the condensate polishers for treatment, minimizing the spread of contamination.

• If leakage occurs from the condensate and feedwater system equipment, the water is drained to local drain hubs by gravity and transferred to the TGB building drain system sump for collection.

• Extraction steam piping is designed to minimize the effects of erosion/corrosion, is adequately sized to limit velocities, and is routed with long-radius elbows to minimize potential erosion and waste generation.

12.4-99 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (50 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 4 Decommissioning

• The SSCs are designed for the full service life and are fabricated as individual assemblies for easy 10.4.7 Planning removal.

• The SSCs are designed with decontamination capabilities. Design features such as welding techniques used and surface finishes are intended to minimize the need for decontamination, and hence reduce waste generation.

• The systems are designed without any embedded or buried piping. The feedwater piping between the TGB and the AB is routed at a high elevation and is provided with a piping sleeve.

Any leakage is drained back to the building and is collected by the floor drain system, thus minimizing unintended contamination of the environment.

5 Operations and

• The components of these systems are located in open areas inside the AB and TGB. Adequate 10.4.7 Documentation spaces are provided around the equipment to enable prompt assessment and responses.

• The COL applicant is to provide operational procedures and maintenance programs as related to leak detection and contamination control (COL 10.4(1)). Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

• The COL applicant is to maintain complete documentation of the system design, construction, design modifications, field changes, and operations (COL 10.4(2)). Documentation requirements are included as a COL information item.

6 Site Radiological

• These systems have the potential to contain low levels of contamination. However, these 10.4.7 Environmental systems are located inside the buildings and drainage is collected and treated before release.

Monitoring Hence, these systems are not required to be directly and individually monitored for environmental contamination.

12.4-100 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (51 of 75)

System: Auxiliary Feedwater System and Auxiliary Feedwater Pump Turbine System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The auxiliary feedwater pump turbine system consists of two non-condensing steam turbines for 10.4.9 of Unintended the auxiliary feedwater pumps and their associated piping and valves. The auxiliary feedwater Contamination pump turbine system is not used during normal power operation, startup, or shutdown. The turbines are either manually or automatically actuated by an auxiliary feedwater actuation signal from the ESFAS or the diverse protection system, which is initiated when the SG water level is low.

• The auxiliary feedwater pump turbine system piping uses stainless steel material and is fabricated to ASME Section III, Class 3 requirements.

2 Adequate and Early Leak

• The valve stem leak-offs and drains are collected and directed to the liquid radwaste system for 10.4.9 Detection treatment and release.

3 Reduction of Cross-

• Normally, the AFWS is not in use and is not radioactively contaminated. There is a segment of 10.4.9 Contamination, pipe, connected to the high-pressure feedwater lines after the containment penetration to the SGs, Decontamination, and that may become contaminated by the feedwater (recycle condensate) up to the containment Waste Generation isolation check valves. This segment of auxiliary feedwater piping is fabricated of stainless steel material, is of welded construction, and is designed to safety Class 3 and seismic Category I requirements. In addition, any leakage developed on this segment of piping is expected to be collected in the sump. Hence, the risk for leakage and the consequence of contamination is low.

This design approach minimizes leakage and unintended contamination of the facility and the environment.

4 Decommissioning

• The SSCs are designed for the full service life and are fabricated as individual assemblies for easy 10.4.9 Planning removal.

5 Operations and

• Complete documentation of system design, construction, design modifications, field changes, and 10.4.9 Documentation operations is to be maintained by the COL applicant. Documentation requirements are included as a COL information item.

12.4-101 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (52 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 6 Site Radiological

• Because of its normal status in standby mode and low possibility for contamination, the potential 10.4.9 Environmental Monitoring for environmental contamination of soil and groundwater from liquid leakage is minimal.

Therefore, inclusion of the auxiliary feedwater system inclusion in the site radiological environmental monitoring program is not required. However, a site radiological environmental monitoring program is included for the whole plant for detection of radiological contamination.

12.4-102 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (53 of 75)

System: Condensate Polishing System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The system design, including all the polishing cation and mixed-bed demineralizer columns, is 10.4.6 of Unintended designed as vendor-packaged modular units with carbon steel material with rubber lining and Contamination welded construction for life-cycle planning, thus minimizing leakage and unintended contamination of the facility and the environment.

• The CPS cation- and mixed-bed demineralizer columns are designed with sufficient capacity to facilitate continuous partial to full flow during normal operation, including anticipated operational occurrences. The columns are designed with periodic resin replacement to prevent buildup of contaminants, thus minimizing the potential of accumulation of significant amount of contamination and waste generation.

• The facility area that houses the system components, including the equipment cubicles that contain radioactively contaminated or potentially contaminated fluid, shall have sloped floors with coating to facilitate the draining of fluid into drain pipes and/or trenches that direct liquid into a local sump. In addition, the coated walls provide smooth surfaces for cleaning and to facilitate liquid draining. To the extent practicable, the cubicles shall also have early leak detection capabilities to detect component leakage, and shall have provisions to initiate alarm signals for operator actions. The facility layout shall facilitate the operators prompt assessment and fast response when needed.

• Sumps that contain contaminated or potentially contaminated fluid shall be equipped with sump tanks and level switches to initiate pumping when sump tank levels reach a predetermined setpoint. The concrete sumps shall be coated and the sump tanks shall have seals to prevent unintended infiltration of liquid into the sumps. The sumps and sump tanks shall be designed to facilitate periodic maintenance and inspection of the coating.

• Piping embedment shall be minimized to the extent practicable. Where embedment cannot be avoided, consideration shall be given to minimizing embedded piping lengths and using double-walled piping with leak detection capabilities on the outer piping.

12.4-103 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (54 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 2 Adequate and Early Leak

• The CPS vessels are designed with pressure instruments to provide reasonable assurance of safe 10.4.6 Detection operation of the SSCs, including the associated piping, and provide alarms to the operating personnel in the event of high pressure due to accumulation of impurities and low pressure from channeling and/or leakages.

• The facility in which the CPS vessels are located is designed with a trench to collect leakage to the sump for processing and disposal. This design minimizes contamination of the facility and the environment from the accumulation of contaminated water.

3 Reduction of Cross-

• The SSCs are designed with life-cycle planning through the use of nuclear industry proven 10.4.6 Contamination, materials compatible with the chemical, physical, and radioactive environment, thus minimizing Decontamination, and waste generation.

Waste Generation

• The condensate is located separately in an area that is shielded and is equipped with a leak collection trench. The trench is sloped and coated to enhance drainage and cleaning to minimize contamination of the facility and the environment. This design approach also minimizes cross-contamination and waste generation.

3 Reduction of Cross-

• The CPS is designed to operate in total, partial, and bypass flow modes to meet secondary 10.4.6 Contamination, chemistry requirements. The system is designed with on-line specific conductivity instruments, Decontamination, and pH, and other process instrumentation for automated operation with manual initiation, and Waste Generation (cont.) sampling and analyses to facilitate plant personnel to determine operating requirements. This design approach minimizes cross-contamination and waste generation.

• Special internals are provided in the cation vessels to allow backwashing of the top layer of cation resin using service air and demineralized water. The process piping containing contaminated solids is properly sized to facilitate flow and maintain sufficient velocities to prevent settling of solids. The piping is designed to reduce fluid traps, thus reducing the decontamination needs and waste generation.

• Utility connections (sluice water and service air) are designed with a minimum of two barriers to prevent cross-contamination.

4 Decommissioning

• The SSCs are designed for the full service life and are fabricated as individual packaged 10.4.6 Planning assemblies for easy removal.

• The CPS is designed with no embedded or buried piping.

12.4-104 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (55 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 5 Operations and

• The CPS polishing operation is designed for automated operations with manual initiation for the 10.4.6 Documentation different modes of operation. Adequate instrumentation, including conductivity, level, flow, and pressure elements, is provided to monitor and control the operations to prevent undue interruption, thus minimizing the spread of contamination and waste generation.

• Operational procedures and maintenance programs are either prepared or to be included as COL Information items. Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

• Complete documentation of design, construction, design modifications, field changes, and operations is to be maintained by the COL Applicant. Documentation requirements are included as a COL 6 Site Radiological

• The CPS is designed with low level of contamination through leakage in the steam generator. 10.4.6 Environmental Monitoring Industry experiences demonstrate that the integrity of the CPS is well maintained, resulting minimal contamination of the facility and the environment. Hence Radiological Environmental Monitoring Program is not required for this system.

12.4-105 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (56 of 75)

System: Condenser Vacuum System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The condenser vacuum system components are located in an open area at the foundation level 10.4.2 of Unintended inside the turbine generator building. The components are designed to be skid mounted or Contamination provided with drip pan and curb. The floors are sloped, coated and provided with drains that are routed to the local drain hubs. This design approach prevents the spread of contamination to the facility and the environment.

• The condenser vacuum system is designed with sufficient capacity and redundancy to support normal operation, including anticipated operational occurrences. The components and piping are fabricated from carbon steel with welded construction for life-cycle planning, as well as minimizing leakage, and unintended contamination of the facility and the environment.

2 Adequate and Early Leak

• The condenser vacuum system is designed with automated operation with manual initiation for 10.4.2 Detection the different modes of operation. Adequate instrumentation, including level and pressure elements and a radiation monitor, is provided to monitor the operations. Upon receipt of a high radiation signal, the vacuum discharge is diverted to the reactor containment building drain sump, and the vent from that sump is processed through the containment ventilation purge system.

This design approach minimizes the spread of contamination and provides early detection.

3 Reduction of Cross-

• The SSCs are designed with life-cycle planning through the use of nuclear industry-proven 10.4.2 Contamination, materials compatible with the chemical, physical, and radiological environment, thus minimizing Decontamination, and waste generation.

Waste Generation

• The condenser vacuum system components are provided with seal water (demineralized water) for decontamination. Chilled water and other utilities are provided to facilitate operations. The utility connections are designed with a minimum of two barriers to prevent the contamination of clean systems.

12.4-106 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (57 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 4 Decommissioning

• The SSCs are designed for the full service life and are fabricated as individual assemblies for easy 10.4.2 Planning removal.

• The SSCs are designed to facilitate decontamination. Design features, such as the welding techniques and surface finishes, are included to minimize the need for decontamination and therefore minimize waste generation.

• The condenser vacuum system is designed without any embedded or buried piping, thus preventing unintended contamination to the environment and facilitating eventual decommissioning.

5 Operations and

• The condenser vacuum system is located in an open area inside the turbine generator building. 10.4.2 Documentation Adequate space is provided around the equipment to enable prompt assessment and responses when required.

• The COL applicant is to establish operational procedures and maintenance programs as related to leak detection and contamination control (COL 10.4 (1)). Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

• The COL applicant is to maintain the complete documentation of system design, construction, design modifications, field changes, and operations (COL 10.4 (2)). Documentation requirements are included as a COL information item.

6 Site Radiological

• The condenser vacuum system is part of the overall plant and is included in the site radiological 10.4.2 Environmental Monitoring environmental monitoring program, for monitoring of the release of non-condensable gases and the potential for environmental contamination. The program includes air sampling and analysis and monitoring of meteorological conditions, hydro-geological parameters, and potential migration pathways of the radioactive contaminants. The site environmental monitoring program is included as a COL information item.

12.4-107 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (58 of 75)

System: Circulating Water System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The CWS, including all the CWS tubes and associated piping, is designed with corrosion- and 10.4.5 of Unintended erosion-resistant materials for life-cycle planning. The tubes are designed with continuous Contamination cleaning features to maintain heat transfer performance and tube integrity, thus minimizing fouling that may result from corrosion.

• The CWS is designed in conjunction with the turbine generator building drain system. The sumps in the turbine building generator building drain system are used to collect above-ground piping leakages and are equipped with dual level instruments to ensure control of liquid level, thus minimizing the spread of contamination and waste generation.

• CW is sampled and analyzed periodically to monitor radiological contamination in the water.

Contamination detected is to be tracked to the source(s) of leakage and corrective actions (repair and/or isolate through tube plugging) are to be initiated. This design approach minimizes leakage and unintended contamination of the facility and the environment.

• The facility that houses the system components, including the equipment cubicles that contain radioactively contaminated or potentially contaminated fluid, shall have sloped floors with coating to facilitate the draining of fluid into drain pipes and/or trenches that direct liquid into a local sump.

• Sumps that contain contaminated, or potentially contaminated fluid shall be equipped with level switches to initiate pumping when sump levels reach a predetermined setpoint. The concrete sumps shall be coated and shall have seals to prevent unintended infiltration of liquid into the concrete sumps. The sumps shall be designed to facilitate periodic maintenance and inspection of the coating.

2 Adequate and Early Leak

• The CWS leakage is detected by sampling and analysis of the circulating water in the piping, and 10.4.5 Detection fluid levels in the DTS sumps, followed by analysis and inspection. This approach is consistent with industry practices and is effective.

3 Reduction of Cross-

• The SSCs are designed with life-cycle planning through the use of nuclear industry-proven 10.4.5 Contamination, materials compatible with the chemical, physical, and radioactive environment, thus minimizing Decontamination, and waste generation.

Waste Generation

• The CWS is designed to operate at a higher pressure than the condenser. This designed approach minimizes infiltration of condensate into the circulating water and is the most effective.

12.4-108 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (59 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 3 (cont.)

• The areas, in which the CWS SSCs are housed, are designed with sloping floors, epoxy coating to provide drainage and cleanable surfaces, and local sumps to collect leakages and overflows.

Cubicle curbs are also provided to reduce cross-contamination and spread of contamination to other areas.

4 Decommissioning

• The CWS is designed with some buried piping in the yard because of the piping size. CW is 10.4.5 Planning sampled and analyzed periodically to monitor radiological contamination in the water.

Contamination detected is to be tracked to the source of leakage and corrective actions (repair and/or isolate through tube plugging) are to be initiated. This design approach minimizes unintended contamination of the soil surrounding the buried pipe.

5 Operations and

• The CWS is designed for automated operations with manual initiation for the different modes of 10.4.5 Documentation operation.

• Adequate ingress and egress spaces are provided for prompt assessments and appropriate responses to tube leaks when and where they are needed.

• The COL applicant is to establish operational procedures and maintenance programs as related to leak detection and contamination control (COL 10.4(1)). Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

• The COL applicant is to maintain complete documentation of the system design, construction, design modifications, field changes, and operations (COL 10.4(2)). Documentation requirements are included as a COL information item.

6 Site Radiological

• The CWS is designed to be a non-radiological system but with the potential of a low level of 10.4.5 Environmental Monitoring contamination through leakage in the condenser tubes. Through monitoring, inspection, and lessons learned from industry experiences, the integrity of the CWS is well maintained, with no noticeable contamination, or to a very low level of contamination of the facility and the environment that is below the detectable limit. Hence the radiological environmental monitoring program is not required for this system.

12.4-109 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (60 of 75)

System: Turbine Generator Building Open Cooling Water System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The TGBOCWS is designed to exchange heat with TGBCCW heat exchangers, and is not to be in 9.2.9 of Unintended close contact with radioactively contaminated components. The TGBOCWS system supplies Contamination cooling water from CW system and returns the water back to the CW discharge header.

• The heat exchangers that the WH System services, are not expected to contain radioactive fluid.

• The heat exchanger seals are designed to leak toward the outside of the heat exchangers and the leaked drains are collected in the TGB floor drain system for treatment and release.

2 Adequate and Early Leak

• Radiation monitoring is not provided for the TGBOCWS system due to the low risk of radioactive Detection contamination. Grab sampling is taken periodically for analysis to confirm that the cooling 9.2.9 water return is not radioactively contaminated.

3 Reduction of Cross-

• The heat exchangers are constructed with titanium material to minimize pinhole leaks. 9.2.9 Contamination,

• The heat exchanger seals are designed to leak toward the outside of the heat exchangers where Decontamination, and leakage is collected in the building sump.

Waste Generation

• A sump is provided for collection of any leakage. The sump is designed with steel liners and is equipped with level instruments to initiate an alarm signal for operator actions.

4 Decommissioning

• The SSCs are designed for the full service life and are fabricated as individual assemblies for easy 9.2.9 Planning removal.

• The TGBOCWS is designed with minimal embedded or buried piping.

12.4-110 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (61 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 5 Operations and

• The TGBOCWS is designed for automated operations with manual initiation for the different 9.2.9 Documentation modes of operation in conjunction with the WT System.

• Adequate ingress and egress spaces are provided for prompt assessments and appropriate responses when and where they are needed.

• The COL applicant is either to prepare or to include operational procedures and maintenance programs (COL 9.2(34)). Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

• Complete documentation of design, construction, design modifications, field changes, and operations is to be maintained by the COL applicant. Documentation requirements are included as a COL information item.

6 Site Radiological

• The TGBOCWS is designed to prevent contamination through leakage in the heat exchangers. 9.2.9 Environmental Monitoring The integrity of the TGBCCW heat exchangers is expected to be well maintained, resulting in no contamination or to a very low level of contamination of the system. Leakage from the system to the facility and the environment is captured by the design. Any residual contamination of the hydrogeology is not likely to be distinguishable from other contamination sources. Hence, the TGBOCWS has low risk and low radiological consequence, and radiological environmental monitoring for the TGBOCWS is not considered effective. 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(34)).

12.4-111 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (62 of 75)

System: Turbine Generator Building Closed Cooling Water System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The TGBCCW system is designed to exchange heat with TGBCCW heat exchangers, and is not 9.2.8 of Unintended to be in close contact with radioactively contaminated components. The TGBCCW system then Contamination supplies cooling water from surge tank to cool TGB equipment (generator hydrogen cooler, ISO phase bus duct cooler, generator stator water cooler, feedwater pump turbine lube oil coolers, etc.)

and returns the water back to the surge tank.

• The equipment heat exchangers that the TGBCCW system services, are not expected to contain radioactive fluid, except the process sampling coolers, which may have low levels of contamination. Due to the infrequent sampling activity, and because the sampling coolers are small, the risk and consequence for WT contamination is considered low.

• The TGBCCW heat exchanger seals are designed to leak towards the outside of the heat exchangers and the leaked drains are collected in the TGB floor drain system for treatment and release.

• Demineralized water is used as makeup to the surge tank. This design approach, working in conjunction with the TGBOCWS, prevents unintended contamination of the WT System.

2 Adequate and Early Leak

• The surge tank is monitored by a pressure and level instruments. A high/low level alarm for the 9.2.8 Detection TGBCCW surge tank is provided in the MCR and RSR for operator actions.

• Radiation monitoring is not provided for the TGBCCW system due to the low risk of radioactive contamination. Grab sampling is taken periodically for analysis to confirm that the cooling water return is not radioactively contaminated.

3 Reduction of Cross-

• A sump is provided for collection of any leakage. The sump is designed with steel liners and is 9.2.8 Contamination, equipped with level instruments to initiate an alarm signal for operator actions.

Decontamination, and

• Boron nitrite is added to control the pH of the system for corrosion inhibition.

Waste Generation

• Nitrogen gas is used to blanket the surge tank to minimize corrosion, thus minimizing waste generation.

4 Decommissioning

• The surge tank is a packaged unit for the full service life and is fabricated as an individual 9.2.8 Planning assembly for easy removal.

• The TGBCCW system is designed with minimal embedded or buried piping.

12.4-112 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (63 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 5 Operations and

• The TGBCCW system is designed for automated operations with manual initiation for the 9.2.8 Documentation different modes of operation.

• Adequate ingress and egress spaces are provided for prompt assessments and appropriate responses when and where they are needed.

• The COL applicant is either to prepare or to include operational procedures and maintenance programs (COL 9.2(34)). Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

• The COL applicant is to maintain complete documentation of system design, construction, design modifications, field changes, and operations (COL 9.2(35)). Documentation requirements are included as a COL information item.

6 Site Radiological

• The TGBCCW system is designed to prevent contamination through leakage in the heat 9.2.8 Environmental Monitoring exchangers. The integrity of the equipment heat exchanger and the TGBCCW heat exchangers is expected to be well maintained through surveillance, resulting in no contamination or to a very low level of contamination of the system. Leakage from the system to the facility and the environment is captured by the design. The TGBCCW surge tank is located at a high elevation.

Leakage from the tank is likely be collected in the drain system inside the TGB. Hence, the TGBCCW system has low risk and low radiological consequence, and radiological environmental monitoring for the TGBCCW is not considered effective. 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(36)).

12.4-113 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (64 of 75)

System: Auxiliary Steam System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The auxiliary steam system condensate receiver tank, vent condenser, and pumps are located in an 10.4.10 of Unintended enclosed area at the foundation level inside the auxiliary building. The boiler is located in its Contamination auxiliary boiling building outside the auxiliary building. The components are designed to be skid-mounted and vendor-packaged units. The floors are sloped, coated and drains are directed to the local drain hubs and then sump. This design approach prevents unintended contamination of the facility and the environment.

• The auxiliary steam system is designed with sufficient capacities and redundancy to support plant operation when main steam is not available. The components are designed in accordance with ASME Section VIII and other applicable codes for life-cycle planning, thus minimizing unintended contamination and waste generation.

• The facility area that houses the system components, including the equipment cubicles that contain radioactively contaminated or potentially contaminated fluid, shall have sloped floors with coating to facilitate the draining of fluid into drain pipes and/or trenches that direct liquid into a local sump. Tank cubicles shall have coated walls to a height that can provide temporary barriers of the contained fluids. In addition, the coated walls provide smooth surfaces for cleaning and to facilitate liquid draining. To the extent practicable, the cubicles shall also have early leak detection capabilities to detect component leakage, overflow, and/or tank rupture, and shall have provisions to initiate alarm signals for operator actions. The facility layout shall facilitate the operators prompt assessment and fast response when needed.

• Sumps that contain contaminated or potentially contaminated fluid shall be equipped with level switches to initiate pumping when sump levels reach a predetermined setpoint. The concrete sumps shall be coated and shall have seals to prevent unintended infiltration of liquid into the concrete sumps. The sumps shall be designed to facilitate periodic maintenance and inspection of the coating.

• Piping embedment shall be minimized to the extent practicable. Where embedment cannot be avoided, consideration shall be given to minimizing embedded piping lengths and using double-walled piping with leak detection capabilities on the outer piping.

• Buried piping in the yard and between buildings and facilities shall be minimized to the extent practicable. Piping tunnels with leak detection capability and accessibility shall be used for piping that contains radioactive or potentially radioactive fluids.

12.4-114 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (65 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 2 Adequate and Early Leak

• The auxiliary steam system is designed with automated operation with manual initiation for the 10.4.10 Detection different modes of operation. Adequate instrumentation, including level and pressure elements, and a radiation monitor, is provided to monitor and control the operations. Upon a high radiation signal, the condensate is diverted to the LWMS for treatment and release. This design approach thus provides early detection and minimizes spread of contamination and waste generation.

3 Reduction of Cross-

• The SSCs are designed with life-cycle planning through the use of nuclear industry-proven 10.4.10 Contamination, materials compatible with the chemical, physical, and radioactive environment, thus minimizing Decontamination, and waste generation.

Waste Generation

• The auxiliary steam system is equipped with a radiation monitor to continuously check the contamination level in the condensate, and sampling and analysis for confirmation and calibration of the radiation monitor, if necessary. If contamination is detected at or above a setpoint, an alarm is initiated for operator actions, and a signal is sent to open the condensate transfer valves to the LWMS for treatment and release. This design approach minimizes cross-contamination to other components.

4 Decommissioning

• The SSCs are designed for the full service life and are fabricated as individual assemblies for easy 10.4.10 Planning removal.

• The SSCs are designed with decontamination capabilities. Demineralized water is provided for makeup as well as for decontamination. Design features such as welding techniques used and surface finishes are intended to minimize the need for decontamination, and hence reduce waste generation.

• The auxiliary steam system is designed with minimum embedded or buried piping, thus minimizing unintended contamination due to leaking of buried or embedded piping. Yard piping is routed in an underground concrete tunnel that is designed with a leakage collection sump, level switch, and pump. An alarm is also provided in the MCR for operator actions in the event of detection of liquid.

12.4-115 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (66 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 5 Operations and

• The auxiliary steam system is designed with adequate instrumentation for automatic operation 10.4.10 Documentation with manual initiation. The boiler is a self-contained vendor package complete with its own instrumentation. The operation of the boiler operation is controlled from a local panel but with remote shutdown at low fuel oil level, or at operator initiation from the MCR.

• The auxiliary steam system condensate receiver tank, the vent condenser, and the pumps are located in an enclosed cubicle inside the auxiliary building, and the boiler is located in an independent auxiliary boiler building for separation purposes. Adequate ingress and egress spaces are provided for prompt assessments and appropriate responses when and where they are needed.

• The COL applicant is to establish operational procedures and maintenance programs as related to leak detection and contamination control in accordance with NRC RG 4.21 (COL 10.4(1)).

• The COL applicant is to maintain the complete documentation of system design, construction, design modifications, field changes, and operations (COL 10.4(2)).

6 Site Radiological

• The auxiliary steam system is designed with a low level of contamination through leakage in the 10.4.10 Environmental Monitoring condensate receiver tank, the vent condenser, pumps, boiler, and the associated piping and instruments. Industry experiences demonstrate that the integrity of the auxiliary steam system is well maintained, and the control methods included in the current design further minimize and/or prevent contamination of the facility and the environment. Hence, a radiological environmental monitoring program is not required for this system.

12.4-116 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (67 of 75)

System: HVAC Systems for Potentially Contaminated Building Areas (Fueling Handling Area HVAC System, Auxiliary Building Controlled Area HVAC System, Reactor Containment Building HVAC System, Reactor Containment Building Purge System, Compound Building HVAC System)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The system components, including ACUs, the reactor cavity AHU, low volume purge supply 9.4.2/9.4.5 of Unintended fans, cubicle coolers, and reactor containment fan coolers (RCFCs) in radiologically controlled 9.4.6/9.4.7 Contamination areas, are fabricated from materials such as carbon and stainless steels that are compatible with the chemical, physical, and radiological environment and are of welded construction for life-cycle planning. The design thus minimizes unintended leakage and unintended contamination of the facility and the environment.

• The low volume purge subsystem of the reactor containment building purge system normally operates in an exhaust mode but can be placed in a recirculation mode. Upon detection of high level of radioactivity at the common discharge duct, the subsystem is placed in the recirculation mode. The discharged air filtered from the low volume purge exhaust ACU flows back into the containment building through the low volume purge supply fan until the level of radioactivity returns back to below the preset value.

• The chilled water side of the reactor cavity AHU, cubicle coolers, and RCFCs is operated at a higher pressure than the atmospheric pressure in the equipment area to prevent the contamination of the chilled water. Corrosion inhibitor chemicals are also added to the chilled water system to minimize the potential of leakage due to corrosion.

• The mounting frames of the HEPA filter and carbon adsorber are stainless steel and of welded construction for life-cycle planning and to prevent unfiltered in-leakage.

• Doors and door frames of all ACUs are the marine bulkhead type or an equivalent airtight construction to minimize leakage and unintended contamination of the facility and the environment.

• All ACU housing penetrations are sealed by welds or adjustable compression-gland type seals to prevent unfiltered in-leakage.

• All drain lines from ACUs, the reactor cavity AHU, cubicle coolers, and RCFCs are sloped and routed to the protective equipment drain. The mounting frames of the high efficient particulate air filters and carbon adsorber are stainless steel and of welded construction for life-cycle planning and to prevent unfiltered leakage.

12.4-117 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (68 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 2 Adequate and Early Leak

• The ACU housing leak test is performed in the manufacturers shop prior to shipment and in the 9.4.2/9.4.5 Detection field following the recommended test frequency described in ASME AG-1 to prevent unfiltered 9.4.6/9.4.7 in-leakage.

• Mounting frame leak tests of HEPA filters and carbon adsorbers are performed in the manufacturers shop prior to shipment and are performed in accordance with ASME AG-1 recommended test frequencies at the site to prevent unfiltered in-leakage.

• Radiation monitors are provided at the outlet of ACUs to monitor the radiological contamination levels to provide reasonable assurance that the effluent release limit will not be exceeded.

3 Reduction of Cross-

• The ACU is designed to operate slightly below atmospheric pressure through the use of an 9.4.2/9.4.5 Contamination, induced fan. This design approach minimizes air leakage from the ACU to its surroundings. 9.4.6/9.4.7 Decontamination, and All ducts are of welded construction to minimize cross-contamination.

Waste Generation

• HVAC component housing surfaces are painted smooth for corrosion protection, ease of decontamination, and cleaning.

• HEPA filters capture radioactive particles and carbon adsorbers remove organic and inorganic forms of iodine from the air stream, thus minimizing the spread of contamination and the resultant waste generation.

• Adequate space is provided around the HVAC equipment to enable prompt assessment and responses when required.

4 Decommissioning

• The ACUs, reactor cavity AHU, cubicle coolers, low volume purge supply fans, and RCFCs are 9.4.2/9.4.5 Planning designed and fabricated as modular units and compartments for easy decommissioning. 9.4.6/9.4.7 12.4-118 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (69 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 5 Operations and

• Adequate work space between mounting frames for the ACUs is provided to facilitate 9.4.2/9.4.5 Documentation maintenance and minimize the potential for the spread of contamination. 9.4.6/9.4.7

• Carbon removal from the carbon adsorber is facilitated by the use of portable pneumatic carbon removal equipment to minimize spillage.

• ACUs, the reactor cavity AHU, cubicle coolers, low volume purge supply fans, and RCFCs are designed with adequate instrumentation to be remotely operated with manual initiation and stopped from the MCR and the RSR.

• The COL applicant is to establish operational procedures and maintenance programs as related to leak detection and contamination control (COL 9.4(4)). Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

6 Site Radiological

• The air quality around the plant is sampled and analyzed routinely for contamination levels and 9.4.2/9.4.5 Environmental Monitoring migration pathways as part of the Site Radiological Environmental Monitoring Program. 9.4.6/9.4.7 Potentially contaminated HVAC systems are expected to be included in this program.

12.4-119 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (70 of 75)

System: Miscellaneous Building Drain System and Turbine Generator Building Drain System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• The DT/DM Systems contain drains that originate from the TGB. These systems consist of 9.3.3 of Unintended sumps and the associated discharge piping for pumping nonradioactive floor and equipment Contamination drainage and condensate overflow to the WWT. The sumps are equipped with sump liners with pumps and level instruments to forward drains for treatment and release.

• The discharge from these pumps is monitored for radiation contamination level. When contamination level is detected at or exceeding a predetermined setpoint, the drains are routed to the LWMS for processing and release. This approach prevent unintended contamination of the WWT.

2 Adequate and Early Leak

• Radiation monitors are provided each of the discharge piping from the TGB north and south pit 9.3.3 Detection sumps and the condensate polishing area sump to detection contamination levels of the drains.

3 Reduction of Cross-

• The sump is designed with stainless steel sump liners that fit into the concrete sump pits. The 9.3.3 Contamination, concrete sump pits is coated to minimize cross-contamination of the TGB basemat and the soil Decontamination, and and groundwater below.

Waste Generation

• The TGB drains are monitored for contamination. Only drains that are radioactively contaminated are routed to LWMS for processing. This design approach minimizes the wastewater to the LWMS, as the wastewater is likely to contain oily waste that can be harmful to the LWMS treatment system. The oily waste is either separated in the DM/DT sumps or in the LWMS normal sump to minimize waste generation.

• The piping for the DM/DT Systems are constructed with stainless stain material and is routed via pipe chase to minimize embedded piping.

4 Decommissioning

• The SSCs are designed for the full service life and are fabricated as individual assemblies for easy 9.3.3 Planning removal.

12.4-120 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (71 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 5 Operations and

• The DM/DT Systems are designed for automated operations with level instruments. The 9.3.3 Documentation radiation monitors are interlocked with the discharge valves to provide the proper routing of the contaminated fluid.

• Adequate ingress and egress spaces are provided for prompt assessments and appropriate responses when and where are needed.

• Operational procedures and maintenance programs are either prepared or to be included as COL Information items. Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

• Complete documentation of design, construction, design modifications, field changes, and operations is to be maintained by the COL Applicant. Documentation requirements are included as a COL Information item.

6 Site Radiological

• Drains and leakage collected in the TGB sumps are generally low in potential that the drain is 9.3.3 Environmental Monitoring radioactively contaminated as condensate and feedwater are continuously polished by ion exchange to keep radiation contamination low. The DM/DT Systems are designed to provide continuous monitoring of drain routing and to prevent spread of contamination. As the contamination level is low, environmental radiation monitoring for the DM/DT Systems is not effective. A site-wide Radiological Environmental Monitoring Program is to be included to monitor the onsite hydrogeology and the potential effects of operation of the plant by the COL Applicant.

12.4-121 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (72 of 75)

System: Radiation Monitoring Systems DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• PERMS use both off-line and in-line type detectors. Process and effluent radiation monitors that 11.5.2 of Unintended come in contact with process fluids are fabricated with stainless steel to prevent unintended Contamination contamination. Readout and alarm units are installed in low-radiation areas.

• The piping is stainless steel material and of welded construction, thus minimizing leakage and unintended contamination of the facility and the environment.

• Area radiation monitors are installed in normally accessible locations. In the event that a detector is located in a high-radiation area, the local readout and alarm unit are located in low-radiation areas for remote readout from the detector.

2 Adequate and Early Leak

• The RMS is designed for automated operation to provide early alarms for process and area 11.5.2 Detection monitoring. Radiation levels and alarms are displayed in the MCR, with the exception of the compound building radiation levels and alarms which are displayed in the radwaste control room.

3 Reduction of Cross-

• The monitors are designed with life-cycle planning through the use of nuclear industry-proven 11.5.2 Contamination, materials compatible with the chemical, physical, and radioactive environment, thus minimizing Decontamination, and waste generation.

Waste Generation

• PERMS monitors are provided with means for the purging or flushing the detector assemblies using nitrogen gas or demineralized water.

• PERMS returns process fluids back to the system to minimize waste generation.

4 Decommissioning

• The SSCs are designed for the full service life and are fabricated as individual assemblies for easy 11.5.2 Planning removal.

• The SSCs are designed with decontamination capabilities. Design features such as welding techniques used and surface finishes are intended to minimize the need for decontamination, and hence reduce waste generation.

• The RMS system is designed with no embedded or buried piping, thus preventing unintended contamination due to the leaking of buried or embedded piping.

12.4-122 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (73 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 5 Operations and

• The RMS system maintains a continuous record of radiation levels for documentation purposes. 11.5.2 Documentation Plant personnel can access all RMS information from the MCR operator console.

• The RMS piping and components are located in the reactor containment building, compound building, and turbine generator building. Adequate ingress and egress spaces are provided for prompt assessment and appropriate responses, when required.

• Operational procedures and maintenance programs are to be prepared by the COL applicant (COL 11.5(8)). Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

• Complete documentation of design, construction, design modifications, field changes, and operations is to be maintained by the COL applicant. Documentation requirements are included as a COL information item.

6 Site Radiological

• The RMS is designed to provide data for plant operation. The RMS components do not process 11.5.2 Environmental Monitoring or generate wastes that can impact contamination of facility and the environment. The RMS is designed to provide data for plant operation and the data can be used to support the radiological environmental monitoring program The program is included as a COL information item (COL 11.5(9)).

12.4-123 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (74 of 75)

System: Containment Monitoring System DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 1 Prevention/Minimization

• All temperature and containment water level instruments are located inside the containment with 6.2.5 of Unintended measurement readout in the MCR for continuous indication of the RCB environment and Contamination containment integrity. The monitoring provides reasonable assurance during normal power operation and minimize the potential for interruption or mishap, thus preventing unintended contamination of the facility and the environment.

• Four channel of pressure instruments are located close to the containment for more accurate measurements. Readouts are provided in the MCR. This design approach minimizes unintended contamination of the facility and the environment.

• Two channels of continuous monitoring of containment atmosphere hydrogen concentration are provided. After the analysis, the air samples are sent back inside the containment.

• The same two channels are also use for continuous monitoring of IRWST hydrogen concentration. After the analysis, the air samples are sent back inside the containment.

2 Adequate and Early Leak

• The CM system is designed to provide continuous indication of containment environment 6.2.5 Detection including pressure, containment water level, hydrogen concentration in the containment and the IRWST, and containment temperature. The instruments monitors the containment integrity for safe operation. The instruments do not process or generate any waste that require leak detection.

3 Reduction of Cross-

• The air samples from the containment and the IRWST using for hydrogen monitoring are returned 6.2.5 Contamination, back to the containment. The air is then process by the low volume purge system. This design Decontamination, and approach minimizes the potential for cross-contamination and waste generation.

Waste Generation

• The temperature and water level instruments are located inside the containment for direct measurement. This design approach also minimizes cross-contamination.

4 Decommissioning

• The instruments are designed for the full service life, including preventive and periodic 6.2.5 Planning replacement of components required for accurate indication.

• The CM system piping is designed with minimal embedded or buried piping.

12.4-124 Rev. 2

APR1400 DCD TIER 2 Table 12.4-10 (75 of 75)

DCD Tier 2 Objective SSC Control Measures / Design Features in DCD to Meet Objective Reference 5 Operations and

• The CM system is designed to provide containment environmental data for monitoring of the 6.2.5 Documentation integrity of the containment. Operational procedures and maintenance programs for these instruments are either prepared or to be included as COL information items. Procedures and maintenance programs are to be completed before fuel is loaded for commissioning.

• Complete documentation of design, construction, design modifications, field changes, and operations is to be maintained by the COL applicant. Documentation requirements are included as a COL information item.

6 Site Radiological

• The CM system is designed to provide continuous monitoring of containment integrity. The 6.2.5 Environmental Monitoring instruments do not process or generate radioactive wastes that contribute to contamination of the facility or the environment. Hence the CM system is in compliance with RG 4.21. A site-wide radiological environmental monitoring program is to be included to monitor the onsite hydrogeology and the potential effects of operation of the plant by the COL applicant.

12.4-125 Rev. 2

APR1400 DCD TIER 2 Drain Hub Leak Detection instrument See Figure 12.4-2 12.3-51 for detailed view Trench or Basemat Buried Double-Walled Valve pit Drain Pipe Drain Header Figure 12.4-1 Typical Design Details of Drainage and Leak Detection Configuration 12.4-126 Rev. 2

APR1400 DCD TIER 2 Radwaste Processor Control Room MCR (Representative alarm)

Main drain pipe in Concrete trench Figure 12.4-2 Typical Design Details of Leak Detection Instrumentation 12.4-127 Rev. 2

APR1400 DCD TIER 2 Figure 12.4-3 Typical Compound Building Trench Location Schematic 12.4-128 Rev. 2

APR1400 DCD TIER 2 Install piping sleeve with flexible joint between building walls Slope sleeve slightly to facilitate drainage Provide caulking on sleeve to seal and flexibility to sleeve expansion Provide sealing material between sleeve and inside pipe to maintain pressure boundary and piping movement Provide drainage Figure 12.4-4 Typical Design for Building Sleeves Schematic 12.4-129 Rev. 2

APR1400 DCD TIER 2 12.5 Operational Radiation Protection Program To achieve the goal of maintaining occupational and public doses both below regulatory limits and as low as is reasonably achievable (ALARA), the operational radiation protection program shall be developed and implemented by the COL applicant in such a way that the program is clearly and sufficiently described in accordance with all applicable guidance (COL 12.1(3)).

The COL applicant is to provide the operational radiation protection program, which includes the following: (Reference 1) (COL 12.5(1))

a Organizational positions, functional responsibilities, experience, and qualifications of personnel responsible for the radiation protection program b Equipment necessary to measure radioactivity as well as radiation fields and exposures including:

1) The number, type, range, sensitivity, calibration method and frequency, availability

2) The planned use of portable, fixed, laboratory, and personnel monitoring instrumentation

c. Health physics facilities and associated protective equipment for controlling ORE and contamination

d. Description of the methods for ensuring development of the training, retraining, and indoctrination program and the radiation protection instruction manuals

e. Procedures:

1) To receive, store, transfer, and dispose of radioactive material

2) To control exposures

3) To control and minimize contamination

4) To facilitate decommissioning

5) To provide adequate radiation monitoring 12.5-1 Rev. 2

APR1400 DCD TIER 2

6) To conduct program reviews and quality assurance

f. Operational programs including:

1) Radiation protection program

2) ALARA program

3) Groundwater protection program

4) Leakage control program 12.5.1 Combined License Information COL 12.5(1) The COL applicant is to provide the operational radiation protection program, including the items described in Section 12.5.

12.5.2 References

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

12.5-2 Rev. 2