ML19254A791
| ML19254A791 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 08/21/1979 |
| From: | Whitmer C GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML19254A790 | List: |
| References | |
| NUDOCS 7908240446 | |
| Download: ML19254A791 (105) | |
Text
{{#Wiki_filter:BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION NRC Docket Nos. 50-424, 50-425 In the Matter of GEORGIA POWER COMPANY SUPPLEMENT 6 TO APPLICATION FOR LICENSE UNDER THE ATOMIC ENERGY ACT OF 1954 AS AMENDED FOR ALVIN W. V0GTLE NUCLEAR PLANT UNITS 1, 2 The Applicant, Georgia Power Company, hereby supplements its Application for a Construction Permit and Opetating License, originally submitted on August 1, 1972, by the addition of supplemen ury material attached hereto.
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~ ' Vice President Sworn to and subscribed before me, this day of August, 1979.
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8 ' hotarv Public Nr.tary Publ.c Georc;a, Stt,te at large My Commission Lapires Sq,t. 29,1981 q Q QOW* ),
en cm , INSTRUCTION SHEET-1 D N SUPPLEMENT NO. 6 ou1
- em cg ALVIN W. VOGTLE NUCLEAR PLANT A PRELIMINARY Lb- - 3 SAFETY ANALYSIS REPORT 00 NOT REM 0W EXISTING WHITE PAGES Replace Table of Contenta pages S2 i and S2 ii with pages S6 i and S6 ii, replace pages S2 xiii and S2 xiv with pages S6 xiii and S6 xiv; replace pages S2 xxiii and S2 xxiv with S6 xxiii ana X6 xxiv.
Replace Table of Contents pages S2 1-1 and S2 1-ii with S6 1-1 and S6 1-ii. Insert Table of Contents page S6 1-vii ahead of page 1-vii. Insert Chapter 1, pages S6 1.1-1 and S6 1.1-2 ahead of page 1.1-1. Insert Chapter 1, pages S6 1.2-9 through 1.2-12 ahead of page 1.2-9. Insert Chapter 1, figures 1.2-9 and 1.2-10 ahead of figure 1.2-9. Insert Table of Contents pages S6 3-xi, S6 3-xia and S6 3-xii ahead of page 3-xi; insert pages S6 3-xix through S6 3-xxii ahead of 3-xix. Insert Chapter 3, pages S6 3.8-81 through S6 3.8-84 ahead of page 3.8-81; insert pages S6 3.8-91 and S6 3.8-92 ahead of nage 3.8-91; insert page S6 3.8-95 and S6 3.8-96 ahead of 3.8-95 and insert page S6 3.8-123 and S6 3.8-124 ahead of page 3.8- 123. Insert figure 3.8-2 ahead of figure 3.8-2, insert figure 3.8-4 ahead of figure 3.8-4. Inscrt figures 3.8-27 and 3.8-28; insert change sheet for figure 3.8-29 and figures 3.8-30 and 3.8-31 ahead of figure 3.8-27. Insert Table of Contents pages S6 6-1 through S6 6-vi ahead of page 6-i; insert page S6 6-xi ahead of page 6-xi. Insert Chapter 6, pages S6 6.1-1 through S6 6.1-2 a/b ahead of page 6.1-1. Incert Chapter 6, page S6 6.2-5 ahead of page 6.2-5; insert page 6.2-51 and 6.2-6 ahead of page 6.2-51. Insert Chapter 6, pages S6 6.6-1 and S6 6.6-6 ahead of page 6.6-1. Insert Chapter 6, Change Sheets for figures 6.6-1 and 6.6-2 ahead of figure 6.6-1. 823 007 DO NOT REMOVE EXISTING WHITE PAGES
INSTRUCTION SHEET-2 DbDl D SUPPLEMENT NO. 6 bg o E< o - - , ALVIN W. VOGTLE NUCLEAR PLANT PRELIMINARY O - -{ U SAFETY ANALYSIS REPORT DO NOT REMOVE EXISTING WHITE PAGES Insert Table of Contents pages S6 7-i and ,S6 7-ii ahead of page 7-1. Insert Chapter 7, page S6 7.3-1 and S6 7.3-2 ahead of page 7.3-1. Insert Table of Contents pages S6 8-i and S6 8-ii ahead of page S-i. Insert Chapter 8, pages S6 8.1-3 and S6 8.1-4 ahead of page 8.1-3. Insert Chapter 8, pages S6 8.3-3 through S6 8.3-Sa in front of page 8.3-3. Insert Table of Contents pages S6 9-vii and S6 9-viii ahead of page 9-vii. Insert Chapter 9, page S6 9.5-11 ahead of page 9.5-11. Insert Table of Contents page S6 12-1 and S6 12-11 ahead of page 12-i. Insert Chapter 12, page S6 12.2-1 and S6 12.2-2 ahead of page 12.2-1. Insert Table of Contents pages S6 15-i through S6 15-iv ahead of '."able of Contents page 15-i. Insert Chapter 15, pages S6 15.4-1 through S6 15.4-10 ahead of r.tge 15.4-1, e Insert Appendix 15B pages S6 ISB-i through S6 15B-5 beh.nd the divider tab for Appendix 15B and ahead of page ISB-i. Insert Table of Contents page S6 16-iii and S6 16-iv ahead of 16-iii; insert page S6 16-xi and S6 16-xii ahead of page 16-xi. I Insert Chapter 16, page S6 16.3.3-3 and S6 16.3.3-4 ahead of page 16.3.3-3. Insert Chapter 16, page S6 16.4.5-3 and S6 16.4.5-4 ahead of page 16.4.5-3. Insert Chapter 16, page S6 16.5.2-1 and S6 16.5.2-2 ahead of page 16.5.2-1. 828 000 00 NOT REM 8VE EXISTING WHITE PAGES
VNP A o JU L S .1 _a TABLE OF CONTENTS Section Title Page I INTRODUCTION AND GENERAL DESCRIPTION 1.1-1 OF PLANT
1.1 INTRODUCTION
1.1-1 1.1.1 LICENSE REQUESTED 1.1-1 1.1.2 PLANT UNITS 1.1-1 1.1.3 PROPOSED PLANT LOCATION 1.1-1 1.1.4 CONTAINMENT TYPE S6 1.1-2 lS6 1..l.4 CONTAINMENT TYPE 1.1-2 1.1.5 NUCLEAR STEAM SUPPLY SYSTEM 1.1-2 1.1.6 SCHEDULE FOR COMPLETION AND COMMERCIAL OPERATION 1.1-3 1.1.7 ORGANIZATION OF CONTENTS l.1-3 1.2
SUMMARY
PLANT DESCRIPTION 1.2-1 1.2.1 SITE CHARACTERISTICS 1,2-1 1.2.2 FACILITY ARRANGEMENT 1.2-5 1.2.3 NUCLEAR STEAM SUPPLY SYSTEM (NSSS) 1.2-6a 1.2.4 STEAM AND POWER CONVERSION SYSTEM 1.2-8 1.2.0 EQUIPMENT BUILDING 36 1.2-9 lS6 1.2.6 ENCLOSURE BUILDING l.2-9 1.2.7 SAFETY FEATURES 1.2-9 1.2.8 UNIT CONTROL 1.2-11 1.2.9 PLANT ELECTRICAL POWER 1.2-11 1.2.10 PLANT INSTRUMENTATION AND CONTROL SYSTEM 1,2-12 1.2.11 AUXILIARY SYSTEMS 1.2-13 9 1.2.12 WASTE PROCESSING SYSTEM 1.2-17 S6 i 828 009 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
N D D VNP bW TABLE OF CONTENTS (Continuc qrg e g n ry g1 F a g Section Title Page 1.2.13 SHARED FACILITIES AND EQUIPMENT 1.2-17 1.3 COMPARISON TABLE 1.3-1 1.3.1 PLANT COMPARISON 1.3-1 1.3.2 COMPARISON OF FINAL AND PRELIMINARY k DESIGN 1.3-1 1.4 IDENTIFICAT1CH OF AGENTS AND CONTRACTORS S2 1.4-1 1.4.1 APPLICANT - CONSTRUCTION MANAGER AND CONTRACTOR S2 1.4-1 1.4.2 SOUTHERN COMPANY SERVICES, INC. S2 1.4-3 1.4.3 BECHTEL POWER CORPORATION S2 1.4-3 1.4.4 NUCLEAR STEAM SUPPLY SYSTEM MANUFACTURER S2 1.4-4 1.4.5 DIVISION OF RESPONSIBILITY S2 1.4-4 1.4 IDENTIFICATION OF AGENTS AND COhrRACTORS 1.4-1 1.4.1 APPLICANT - GENERAL CONTRACTOR AND OPERATOR 1.4-1 1.4.2 ARCHITECT - ENGINEER 1.4-3 1.4.3 NUCLEAR STEAM SUPPLY SYSTEM FANUFACTURER 1.4-4 1.4.4 DIVISION OF RESPONSIBILITY 1.4-4 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.5-1 1.5.1 CORE STABILITY EVALUATION PROGRAM 1.5-1 1.5.2 FUEL DEVELOPMENT PROGRAM FOR OPERATION AT .iIGH POWER DENSITIES 1.5-1 1.5.3 IN-CORE DETECTOR PROGRAM 1.5-2 028 010 1 S6 ii POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATI N -
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VNP D 1[ A s.3 b- -] TABLE OF CONTENTS (Continued) Section Title Page 5.6 INSTRUMENTATION APPLICATION 5.6-1 6 ENGINEERED SAFETY FEATURES 6.1-1 6.1 GENERAL 6.1-1 6.1.1 SAFETY FEATURES SYSTEMS S6 6.1-1 lS6 6.1.1 SAFETY FEATURES SYSTEMS 6.1-1 6.1.2 OPERATIONAL RELIABILITY 6.1-2 6.2 CONTAINMENT SYSTMiS 6.2-1 6.2.1 CONTAINMENT FUNCTIONAL DESIGN 6.2-1 6.2.2 CONTAINMENT HEAT REMOVAL SYSTEMS 6.2-33 6.2.3 CONTAINMENT AZ_". PURIFICATION AND CLEANUP SYSTEMS 6.2-53 6.2.4 CONTAINMENT ISOLATION SYSTEMS 6.2-76 6.2.5 COMBUSTIBLE GAS CONTROL IN CONTAINMENT 6.2-91 6.3 EMERGENCY CORE COOLING SYSTEM 6.3-1 6.4 NUCLEAR SERVICE WATER AND COMPONENT COOLING WATER SYSTEMS 6.4-1 6.4.1 NUCLEAR SERVICE COOLING WATEx SYSTEM 6.4-2 6.4.2 COMPONENT COOLING WATER SYSTEMS 6.4-9 6.4.3 ULTIMATE 11 EAT SINK 6.4-15 6.4.4 SYSTEM EVALUA2 ION 6.4-30 6.4.5 FAILURE MODE ANALYSIS 6.4-34 6.5 PENETRATION ROOM FILTRATION SYSTEMS 6.5-1 6.5.1 DESIGN BASES 6.5-1 6.5.2 SYSTEM DE3I.,N 6.5-la 6.5.3 SYSTEM DESIGN EVALUATION 6.5-5 6.5.4 TESTS AND INSPECTIONS , 6.5-9 S6 xiii c- g 1 -
I $D d D g US LL VNP QfD A TABLE OF CONTENTS (Continued) Section Title Page 6.5.5 INSTRUMENTATION APPLICATION 6.5-13 6.
5.6 REFERENCES
6.5-14 S6l 6.6 DELETED S6 6.6-1 6.6 ENCLOSURE BUILDING FILTRATION AND VENT SYSTE*' (EBFVS) 6.6-1 6.6.1 DESIGN BASES G.6-1 6.6.2 SYSTEM DESIGN 6.6-la 6.6.3 DESIGN EVALUATION 6.6-6 6.6.4 TESTS AND INSPECTIONS 6.6-8 6.6. INSTRUMENTATION APPLICATION 6.6-12 6.
6.6 REFERENCES
6.6 6A IODINE REMOVAL EFFECTIVENESS EVALUA-TION OF CONTAINMENT SPRAY SYSTEM 6A-1 6B MATERIALS COMPATIBILITY REVIEW 6B-1 7 INSTRUMENTATION AND CONTROLS 7.1-1
7.1 INTRODUCTION
7.1-1 7.2 REACTOR 2 RIP SYSTEM 7.2-1 7.3 ENGINEERED FAFETY FEATURES ACTUATION SYSTEM 7.3-1 7.3.4 MANUALLY-ACTUATED E*1GINEERED SAFETY FEATURES SYSTEM EQUIPMENT 7.3-4 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 7.4-1 7.4.2 ANALYSIS 7.4-1 7.5 SAFETY RELATED DISPLAY INSTRUMENTATION 7.5-1 7.5.2 ANALYSES 7.5-1 7.5.3 SAFETY FEATURES EQUIPMENT BYPASS INDICATION 7.5-6 828 012 S6 xiv PCST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
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TABLE OF CONTENTS (Con'inued) c Section Title Page 15.4.6 ENVIRONMENTAL CONSEQUENCES OF A POSTULATED ROD EJECTION ACCIDENT 15.4-24 15.
4.7 REFERENCES
15.4-29 15A HEAT TRANSFER COEFFICIENTS USED IN THE LOCTA-R2 CORE THERMAL ANALYSIS 15A-1 ISB DOSE MODELS USED TO EVALUATE THE S6 ENVIRONMEN'2AL CONSEQUENCES OF ACCIDENTS S6 15B-1 ISB DOSE MODELS USED TO F'ALUATE THE ENVIRONMENTAL CONSEQU4NCES OF ACCIDENTS 15B-1 16 TECHNICAL SPECIFICATIONS 16.1 1 DEFINITIONS 16.1.1-1 1.1 POWER LEVEL 16.1.1-1 1.2 REACTOR OPERATING CONDITIONS 16.1.1-1 1.3 OPERABLE 16.1.1-2 1.4 PROTECTIVE INSTRUMENTATION LOGIC 16.1.1-2 1.5 DEGREE OF REDUNDANCY 16.1.1-2 1.6 INSTRUMENTATION 16.1.1-3 1.7 CONTAINMENT INTEGRITY 16.1.1-3 1.8 ABNORMAL OCCURRENCE 16.1.1-4 1.10 QUADRANT POWER TILT 16.1.1-5A 1.11 SAFETY LIMITS 16.1.1-6 1.12 LIMITING SAFETY SYSTEM SETTINGS 16.1.1-6 1.13 LIMITING CONDITIONS FOR OPERATION 16.1.1-6 1.14 SURVEILLANCE REQUIREMEN.'S 16.1.1-6 1.15 LOW POWER PHYSICS TESTS 16.1.1-6 823 013 S6 xxiii ( - - 7
TABLE OF CONT TS (Continued) i L Section Title Page 2 CAFETY LIMITS AND LIMITING 9AFETY SYSTEM SETTINGS 16.2.1-1 2.1 SAFETY LIMIT, REACTOR CORE 16.2.1-1 2.2 SAFETY LIMIT, REACTOR COOLANT SYSTEM PRESSURE 16.2.2-1 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION 16.2.3-1 3 LIMITING CONDITIONS FOR OPERATION 16.3.i 1 3.1 REACTOR COOLANT SYSTEM 16.3.1-1 3.2 CHEMICAL AND VOLUME CONTROL SYSTEM 16.3.2-1 3.3 ENGINEEREO SAFETY FEATURES 16.3.3-1 3.4 SECONDARY STEAM AND POWER CONVERSION SYSTEM 16.3.4-1 3.5 INSTRUMENTATION SYSTEM - OPERATIONAL SAFETY INSTRUMENTATION 16.3.5-1 3.6 CONTAINMENT SYSTEMS 16.3.6-1 3.7 AUXILIARY ELECTRICAL SYSTEMS 16.3.7-1 3.8 REFUELING 16.3.8-1 3.9 RADIOACTIVITY EFFLUENT RELEASE 16.3.9-1 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS 16.3.10-1 3.11 CORE SURVEILLANCE INSTRUMENTATION 16.3.11-1 3.12 LOOP STOP VM VE OPERATION 16.3.12-1 3.13 NTAINMENT POLAR CRANE 16.3.13-1 4 SURVEILLANCE REQUIREMENTS 16.4.1-1 4.1 OPERATIONAL SAFETY REVIEW 16.4.1-1 4.2 PRIMARY SYSTEM SURVEILLANE 16.4.2-1 828 014 S6 xxiv u e _ e
D l 0} mucJJb VNP - - -, TABLE OF CONTENTS gU M_b]_[ g CHAPTER ~l Section Title Page 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 1.1-1
1.1 INTRODUCTION
1.1-1 1.1.1 LICENSE REQUESTED 1.1-1 1.1.' PLANT UNITS 1.1-1 1.1.3 PROPC ED PLANT LOCATION 1.1-1 1.1.4 CONTAINMENT TYPE S6 1.1-2 l36 1.1.4 CONTAINMENT TYPE 1.1-2 1.1.5 NUCLEAR STEAM SUPPLY SYSTEM 1.1-2 1.1.5.1 Reactor Type and Supplier 1.1-2 1.1.5.2 Power Qttput 1.1-2 1.1.6 SCHEDULE FOR COMPLETION AND COMMERCIAL OPERATION 1.1-3 1.1.7 ORG \NIZATION OF CONTENTS 1.1-3 1.1.7.1 Subdivisions 1.1-3 w 1.1.7.2 Standard Format 1.1-3 1.1.7.3 References 1.1-4 1.1.7.4 Talles and Figures 1.1-4 1.1.7.5 Numbering of Pages 1.1-4 1.1.7.6 Amending the PSAR 1.1-4 1:; 2
SUMMARY
PLANT DESCRIPTION 1.2-1 1.2.1 SITE CHARACTERISTICS 1,2-1 1.2.1.1 Location 1.2-1 1.2.1.2 Site Ownership 1.2-1 1 828 015 S6 1-i POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFOR!IATION - AUGUST 21 1979
VNP TABLE OF CONTENTS (Continued) Section Title Page 1.2.1.3 Access to the Site D 1.2-1 1.2.1.4 Site Environs 1.2-1 1.2.1.5 Geology U UU ' 1.2-3 1.2.1.6 Seismology 1.2-3 1.2.1.7 Hydrology 1.2-4 1.2.1.8 Met 3orology 1.2-4 1.2.1.9 Environmental Radiation Monitoring 1.2-4 1.2.2 FACILITY ARRANGEMENT 1.2-5 1.2.2.1 General Arrangement 1.2-5 1.2.2.2 Criterion for Plant Layout 1.2-6 1.2.3 NUCLEAR STEAM SUPPLY SYSTEM (NSSS) 1.2-6a 1.2.3.1 Reactor Core 1.2-7 1.2.3.2 Reactor Coolant System 1.2-7 1.2.4 STEAM AND POWER CONVERSION SYSTEM 1.2-8 1.2.5 CONTAINMENT 1.2-8 S6l 1.2.6 EQUIPMENT BUILDING S6 1.2-9 1.2.6 ENCLOSURE BUILDING 1.2-9 1.2.7 SAFETY FEATURES 1.2-9 1.2.7.1 Emergency Core Cooling System 1,2-9 1.2.7.2 Containment Spray System 1.2-10 1.2.7.3 Containment Cooling System 1.2-10 1.2.7.4 Penetration Room Filtration System 1.2-10 S6l 1.2.7.5 Deleted S6 1.2-10 1.2.7.5 Enclosure Building Filtration System 1.2-10 828 016 S6 1-ii POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
WP O O d6 LIST OF FIGURES A O . - O Figure Title 1.2-1 General Vicinity Map 1.2-2 Site Plot Plan Sheec 1 of 2 1.2-2 Site Plot Plan Sheet 2 of 2 1.2-3 General Arrangement Plan - Elev. 260' - 0" 1.2-4 General Arrangement Plan - Elev. 240' - 0" 1.2-5 General Arrangement Plan - Elev. 220' - 0" 1.2-6 General Arrangement Plan - Elev. 200' - 0" 1.2-7 General Arrangement Plan - Elev. 180' - 0" 1.2-8 General Arrangement Plan - Elev. 140' - 0", Sheet 1 of 2 1.2-8 General Arrangement Plan - Elev. 160' - 0", Sheet 2 of 2 1.2-9 General Arrangement Sect. Elevation Looking West S6 S6 1.2-10 General Arrangement Sect. Elevation Looking North S6 1,2-9 General Arrangement Sect. Elevation Looking North 1.2-10 General Arrangement Sect. Elevation.Looking West 1.2-11 Plot Plan General Arrangement of Equipment 828 Oi7 S6 1-vii POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21 1979
N n> D id l CV cv 5 VNP g3 g- - o . . _-) C1 PTER 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 1.1 .ITRODUCTION This Preliminary Safety Analysis Report (PSAR) complies with the " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants" issued by the Atomic Energy Commission in February 1972. For a discussion of the format of this report, refer to subsection 1.1.7, Organization of Contents. 1.1.1 LICENSE REQUESTED This PSAR is submitted to support the application of the Georgia Power Company, herein referred to as the GPC, for a construction permit and operating license for a nuclear power plant designated as the Alvin W. Vogtle Nuclear Plant, herein referred to as the VNP. 1.1.2 PLANT UNITS The application is for four units, each with a reactor core rated at a power level of 3411 MWt under section 103 of the Atomic Energy Act of 1954, as amended, and the regulations of the Atomic Energy Commission set forth in Part 50 of Title 10 of the Code of F3deral Regulations (10 CFR 50). The plant will be constructed in 2 two-unit stations with the two units in each station essentially the same. Descriptions of one unit shall be interpreted as applying to both units in each station. Similarly, descriptions of one station shall be interpreted as applying to both stations. Differences between the two units in a statica and particularly structures, systems, and components that are shared between the two units are specified in the appropriate location in the PSAR. Similarly, differences between the two stations and any structures, systems, and components shared between the two stations, if any, are likewise specified. 1.1.3 PROPOSED PLANT LOCATION The proposed location of the VNP is on the southwest side of the Savannah River approximately 23 river miles upstream from the intersection of the Savannah River and U.S. Highway
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LU vid S6 1.1-1 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFOPJ1ATION - AUGUST 21, 1979
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the eastern sector of Burke County, Georgia, and across the river from Barnwell County, South Carolina. The VNP site is directly across the' Savannah River from the AEC's Savannah River Plant (SRP). 1.1.4 CONTAINMENT TYPE The cuntainment for each of the VNP units will be designed by Bechtel Corporation, consisting of a prestressed concrete, S o, steel-lined, cylindrical structure with hemispherical dome, called the containment, which completely encloses the reactor coolant pressure boundary. 1.1.5 NUCLEAR STEAM SUPPLY SYSTEM 1.1.5.1 Reactor Type and Supplier The nuclear steam supply system (NSSS) for each of the four VNP units is a pressurized water reactor (PWR). The Westinghouse Electric Corporation is contracted to design and supply units for the VNP. 1.1.5.2 Power Output Each NSSS unit is rated or guaranteed at a net core power out-put of 3411 MWt plus 14 MWt net of heat from nonreactor O sources, primarily pump heat, a total of 3425 MWt; the corresponding turbine-generator gross generator output is 1159 MWe. The design net core power output rating for each NSSD unit is 3565 MWt, he ultimate expected capability of the NSSS. The turbine-generator unit of the steam and power conversion system will have the capability of generating a gross electrical output of 1210 MWe, the maximum calculated-not guaranteed, valves wide open condition, corresponding to a NSSS output of 3570 MWt. Although the license application is for 3411 MWt net core per NSSS unit, all safety systems, including the containment and engineered safety features, are designed and evaluated for operation at a higher power level, 3565 MWt net core. This power rating is used in the analysis of all postulated accidents bearing significantly or the acceptability of the site. The thermal-hydraulic and nuclear aspects of the core have been evaluated on the basis of a core thermal output of 3411 MWt. O 823 019 S6 1.1-2 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
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'V 4 d .L 13 1.2.6 EQUIPMENT BUILDING The equipment building (EB) surrounds the containment from grad ^ to the 270 foot level. The EB is a group of steel-frat d structures with uninsulated metal siding and metal S6 roof deck. It provides protection from the weather for equip-ment located within the building. Safety-related equipt'nt located within the EB are physically protected from tornaao missiles on an individual basis. A detailed description of the structure is provided in section 3.3.4.1.1.
1.2.7 SAFETY FEATURES The safety features limit the potential radiation exposure to the public and to plant personnel follow;.ng an accidental release of radioactive fission products trom the reactor system, particularly as the result of a LOCA. These safety features func ion to localize, control, mitigate, and terminate such accidents, ensuring that 10 CFR 100 limits are not exceeded. The safety features consist of the following systems: A. Emergency core cooling system B. Containment spray system C. Containment cooling system D. Penetration room filtration system E. Hydrogen recombiners and emergency purge system 1.2.7.1 Emergency Core Cooling System The emergency core cooling system (ECCS) injects borated water into the reactor coolant system following a LOCA. This 823 020 S6 1.2-9 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP provides cooling to limit core damage, metal-water reactions, and fission product release; and assures adequate shutdown margin regardless of temperature. The ECCS also provides continuous long term post-accident cooling of the core by recirculating borated water between the containment sump and the reactor core. 1.2.7.2 Containment Spray System The containment spray system is one of two independent, full capacity systems with which each unit is equipped for cooling the containment atmosphere after the postulated LOCA. The containment spray system supplies borated water to cool the containment atmosphere. The spray system in comt' nation with two of the four containment air coolers (operating at reduced speed) is sized to provide adequate cooling w.t.th either or both of the two containment spray pumps in service. These pumps take suction from the refueling water storage tank. When the supply from the storage tank is depleted, suction of the pumps is aligned to pump water from the containment sump directly into the containment during the recirculation mode of operation. Sodium hydroxide is added to the spray to remove iodine from the containment atmos;aere in the post-accident condition. 1.2.7.3 Containment Cooling System The containment cooling system is the second of two independent, full capacity systems with which each unit is equipped to mix and cool the containment atmosphere. 1.2.7.4 , Penetration Room Filtratio_n System The penetration room filtration system for each unit collects and processes potential containment penetration leakage to limit environmental act.ivity levels following a LOCA and the subsequent pressure transient inside the containment. S6l 1.2.7.5 Deleted 0Oh g aMdGM e 828 021 S6 1.2-10 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP lS6 1.2.7.6 Hydrogen Recombiners and Emergency Purge System Electrical hydrogen recombiners reduce the percentage of hydrogen in the post-LOCA containment atmosphere to below uncontrolled release mixture levels. Emergency purge of the containment supplements the recombiners and further reduces hydrogen concentration by feeding ambient air and bleeding containment atmosphere. 1.2.8 UNIT CONTROL The reactor is controlled by control rod aovement and regulation of the boric acid concentration in the reactor coolant. During steady-state operation, the reactor control system maintains a programmed average reactor coolant temperature that rises in proportion to the load. The ceLbined actions of the control system, and steam bypass to the condenser maintain and strengthen steam 'elief station auxiliary load upon separation from the ;ransmission system from full load. The solid state protection logic system automatically initiates appropriate action whenever the parameters monitored by this system reach pre-established setpoints. This system acts to trip the reactor, actuate emergency core cooling, close containment isolation valvas, and initiate the operaticn of other safety features systems. 1.2.9 PLANT ELECTRICAL POWER The four main turbine-generators are each rated at 1350 MVA, 0.90 pf, 22,000 volts; they are 3-phase, 60-Hz, 1800-rpm hydrogen- and water-cooled units. The power from these units is delivered to the GPC 230-kV and 500-kV switchyards, which are connected to the system grid by 230-kV and 500-kV transmission lines. Termination points of these lines are shown in section 8.2. Each unit has three separate sources of power for its auxiliaries. The three sources of power and associated electrical equipment ensure the functioning of all units without undue risk to the health and safety of the public.
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VNP The three sources for each unit are as follows: A. The main turbine-generator supplies normal auxiliary loads during plant operation. B. The two reserve auxiliary transformers supply the safety feature buses from the 230-kV switchyard. C. The two, fast-starting diesel generators are connected to two safety feamrc cuses. Upon loss of all offsite power, either diesel generator and its associated bus can shut the unit down safely under LOCA conditions. Station batteries ensure a constant supply of power to vital instruments and controls. Station power is distributed through redundant buses at 4160 , and 480-volt, ac, to buses and load centers. The de load requirements for each unit are distributed through two 125-volt buses. 1.2.10 PLANT INSTRUMENTATION AND CONTROL SYSTEM To avoid undue risk to the health and safety of the public, instrumentation and controls monitor and maintain neutron flux, primary coolant pressure, temperature, and control rod positions within prescribed operating ranges. The non-nuclear regulating, process, and containment instrumenta-tion measures temperatures, pressure, flow, and levels in the steam systems, containment, and auxiliary systems. Process variables required on a continuous basis for startup, operation, and shutdown of the unit are indicated, recorded, and controlled from the control room. The quality and types of process instrumentation provided ensure safe and orderly operation of all systems and processes over the full operating range of the plant. Startup and shutdown of the reactor and adjustment of reactor power in response to turbine load demand, are provided by the reactor control system. The reactor is controlled by a combination of mechanically-driven control rods. The control system permits each unit to accept step load increases or decreases of 10 percent and ramp load increases or decreases of 5 percent per minute over the load range from 15 percent to, but not exceeding, 100 percent "ower under normal operating conditions, subject to xenon limitations. D G IP M o O O ~9~I A w - - 2 S6 1.2-12 POST-CONSTRUCTIO'; DERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
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JL VNP , _ uu _u _ TABLE OF CONTENTS (Continued) UJy_)]_]( g Section Title Page 3.8.3.6 Materials Quality Control and Special Construction Techniques 3.8-79 3.8.3.7 Testing and Inservice Surveillance Requirements 3.8-81 3.8.4 OTHER CATEGORY I STRUCTURES 3.8-82 3.8.4.1 Description of the Structure S6 3.8-82 lS6 3.8.4.1 Description of the Structure 3.8-82 3.8.4.2 Applicable Codes, Standards and Specifications 3.8-86 3.8.4.3 Loads and Loading Combinations 3.8-88 3.8.4.4 Design and Analysis Procedures S6 3.8-97 lS6 3.8.4.4 Design and Analysis Procedures 3.8-92' 3.8.4.5 Structural Acceptance Criteria 3.8-114 3.8.4.6 Materials, Quality Control, and Special Construction Techniques 3.8-116 3.8.4.7 Testing and Inservice Surveillance Requirements 3.8-117 3.8.5 FOUNDATIONS AND CONCRETE STRUCTURES 3.8-117 3.8.5.1 Description of the Foundations and S6 Supports S6 3.8-117 3.8.5.1 Description of the Foundations and Supports 3.8-117 3.8.5.2 Applicable Codes, Standards and Specifications 3.8-123 3.8.5.3 Loads and Loading Conditions 3.8-123 3.8.5.4 Design and Analysis Procedures 3.8-123 3.8.5.5 Structural Acceptance Criteria 3.8-124 823 030 S6 3-xi e -
VNP TABLE OF CONTENTS (Continued) Section Title Page 3.8.5.6 Materials Quality control and Special Construction Techniques 3.8-124 3.8.5.7 Testing and Inservice Surveillance Requirements 3.8-124 0 0 oo _ I. f A a 823 031 S6 3-xia lS6 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP I O TABLE OF CONTENTS (Continued) {U! Section Title Page 3.9 MECHANICAL SYSTEMS AND COMPONENTS 3.9-1 3.9.1 DYNAMIC SYSTEM ANALYSIS AND TESTING 3.9-1
.9.1.1 Dynamic System Analysis and Testing of Components (and Vibration Operational Test Program) 3.9-1 3.9.1.2 Operability Assurance Program 3.9-1 3.9.1.3 Dynamic Analysis of Reactor Internals 3.9-7a 3.9.1.4 Preoperational Tests 3.9-7a 3.9.1.5 Dynamic Analysis of Safety Related Mechanical Equipment 3.9-8 3.9.1.6 Inelastic Stress Analysis 3.9-8 3.9.1.7 Core components 3.9-8 3.9.2 ASME CODES CLASS 2 AND 3 COMPONENTS 3.9-9 3.9.2.1 Design Pressure and Temperature 3.9-9 3.9.2.2 Design Loading Combinations 3.9-9 3.9.2.3 Stress Limits Associated with Design Loading Combinations 3.9-9 3.9.2.4 Stress Limits Result in Inelastic -
Deformation 3.9-19 3.9.2.5 Code Interpretations 3.9-19 3.9.2.6 Active Pumps and Valves 3.9-19 3.9.2.7 Design Criteria for Pipe Failure 3.9-19 3.9.2.8 Pressure Relieving Devices Design Criteria 3.9-20 3.9.2.9 Design Methods for Safety-Related FTe'chanica 3 Equipment 3.9-25 O B2B 032 S6 3-xii POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP TABLE OF CONTENTS (Continued) D Ch DI ) u 0 JL
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LIST OF FIGURES D '( oS.U.k _a Figure Title 3.7-25 1% Damping Floor Response Spectrum Curve at Mass Pt. 11 of the Containment Structure 3.7-26 Illustrations of the use of Floor Spectrum Curve when more than one Equipment Frequency is within the Widened Spectral Peak 3.7-27 Position of the Structure when overturning about one Edge 3.7-28 Idealized Passive Soil Pressure for Over-turning about Edge R 3.7-29 Buoyancy Effects for Overturning Evaluation 3.7-30 Soil Pressure Prediction 3.8-1 Interior Structure Plans 3.8-2 Interior Structure ?ypica. cactions S6 l56 3.8-2 Interior Structure Typical Sections 3.8-3 Interior Structure Section B 3.8-4 Containment General Arrangement Typical S6 Details S6 3.8-4 Containment General Arrangement Typical Details 3.8-5 Prastressing Tendons Layout and Details 3.8-6 Typical Reinforcement and Tendon Arrangement 3.8-7 Reinforcement and Tendon Arrangement it Major Openings 3.8-8 Major Openings Sections and Details 3.8-9 Liner Plate and Stiffeners 3.8-10 Containment Penetrations-Details 3.8-11 Temperatures 828 ; 3 S6 3-xix POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP TABLE OF CONTENTS (Continued) LIST OF FIGURES Figure Title h D 3.8-12 Design Thermal Gradient Across Cont Wall Units 1 and 2 3.8-13 Model of Containment and Interior Structure for ASHSD Computer Program 3.8-14 Containment Stress Analysis Dead Load 3.8-15 Containment Stress Analysis Prestress Load 3.8-16 Containment Stress Analysis Thermal Operating Load T=2630 F 3.8-18 Containment Stress Analysis Active Soil Pressure Load PA=0.034h KSF 3.8-19 Containment Stress Analysis Passive Soil Pressure Load Pp=020h KSF 3.8-20 Containment Stress Analysis Polar Crane Load 3.8-21 Containment Stress Analysis Pressure Load P-52 Psig 3.E-22 Containment Stress Analysis Horizontal RMS Seismic 0.129 at 2% 3.8-23 Containment Stress Analysis Vertical RMS Seismic 2/3 x 0.12g at 2% 3.8-24 Containment Stress Analysis Horizontal RMS Seismic 0.2g at 5% 3.8-25 Containment Stress Analysis Vertical RMS Seismic 2/3 x 0.2g at 5% 3.8-26 Containment Stress Analysis Tornado Load V=300 MPH 9 1828 034 S6 3-xx POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VN? mG D TABLE OF CONTENTS (Continued) o W 11 ' T'o'9~T A LIST OF FIGURES f y (0] _ 2 Figure Title 3.8-27 Equipment Building Plan at Finish Grade S6 3.8-28 Equipment Building Roof Framing Plans S6 3.8-29 Deleted S6 S6 3.8-30 Equipment Building Sectional Elevations S6 3.8-31 Equipment Building Elevations S6 g 3.8-27 Enclosure Building at Finish Grade 3.8-28 Enclosure Building Roof Framing Plans 3.8-29 Enclosure Building Intermediate Horizontal Framing 3.8-30 Enclosure Building Sectional Elevations 3.8-31 Enclosure Building Elevations 3.8-32 Auxiliary Building Plan at Elevation 220'-0" 3.8-33 Auxiliary Building Plan at Elevation 140'-0" 3.8-34 Cross Section Through Auxiliary Euilding ,. 3.8-35 Control Building Plan at Elevation 220'-0" 3.8-36 Control Building Plan at Elevation 180'-0" 3.8-37 Control Building Section Through Building 3.8-38 Auxiliary and Control Building 3.8-39 Fuel Handling Building Plan at Elevation 220'-0" 3.8-40 Fuel Handling Building Section Through Building 3.8-41 Nuclear Service Cooling Towers S6 3-xxi 823 037) . POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFOPMATION - AUGUST 21, 1979
VNP TABLE OF CONTENTS (Continued) LIST OF FIGURES Figure Title 3.8-42 Finite Element Model of Cooling Tower for Dead Loade Live Loads and Thermal Loads (Sheet 1)
'. 8-42 Mathematical Model Cooling Towers (Sheet 2) 3.8-43 Finite Element Model for Determination of Soil Spring Constants
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VNP TABLE OF CONTENTS (Continued) LIST OF FIGURES Figure Title 3.8-44 Finite Element Model of Cooling Tower for Seismic Analysis 3.8-45 Cooling Tower Stress Analysis Structure and Equipment Weight 3.8-46 Cooling Tower Stress Analysis Liquid Pressure 3.8-47 Cooling Tower Stress Analysis Earth Pressure at Rest 3.8-48 Cooling Tower Stress Analysis Live Load 3.8-49 Cooling Tower Stress Analyeis Operating Thermal Load 3.8-50 Cooling Tower Stress Analysis w cident Thermal Load 3.8-51 Cooling Tower Stress Analysis Earthquake-Hori;:ontal Load O 3.8-52 Cooling Tower Stress Analysis Design Base Earthquake Vertical Load 3.8-53 Localtions of Reference Sections Cooling Towers 3.8-54 Category I Reinforced Concrete Tanks 3.8-55 NSCW and Miscellaneous Categorv I Structurcs Key Plan 3.8-5r nCCW and Miscellaneous Category I Structures Sections 3.8-57 Long Term Effect Creep and Shrinkage 3.8-58 Polar Crane Seismic Retainer Brackec 3.5-59 VSL and Prescon Tendon End Anchors D i \
%a e9 828 037 S6 3-xxii POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP 2.8.3.6,5 ,truction Pro';edure The construction procedures are the same as described in paragraph 3.8.1.6.6. 3.8.3.6.6 Quality Control The quality control requirements will be met as described in paragraph 3.8.1.6.7 and Chapter 17. 3.8.3.7 Testing and Inservice Surveillance Requirements A formal program of testing and inservice surveillance is not planned for the internal strvetures. The internal structures are not directly related to the functioning of the containment concept, hence, no testing or surveillance is required. D
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F e JO ulbh A 3 828 038 S6 3.8-81 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP 3.8.4 OTHER CATEGORY I STRUCTURES S6l Other Category I structures include equipment building, auxiliary building, control building, fuel handling building, diesel generator building, auxiliary feedwater pump building, nuclear service cooling tower, emergency cooling water wells, condensate storage water tank, reactor makeup storage tank, refueling water storage tank, diesel oil fuel storage tank, pipe and electrical cable tunnels and electrical duct banks. 3.8.4.1 Description of the Structure 3.8.4.1.1 Equipment Building The equipment building (EB) is a group of five steel-framed structures with.pninsulated metal siding and roof deck. The EB corapletely surrounds the containment from grade to 270 foot level. There is no safety function associated with the S6 EB except that its failure shall not damage any safety-related equipment. The approximate center line to center line column dimensions of the EB are: length, 176 feet, width, 176 feet, and height, 50 feet above grade. O O D k i h ( b h O na # e S6 3.8-82 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP This space has been left blank intentionally lS6 3.8.4.1.2 Auxiliary Building The auxiliary building is a multistory reinforced concrete structure located south of the fuel handling building with a building separation in between. This building has two wing structures (one east wing and one vest wing) with a separation in between the wings and a combined "oundation mat. It has a four-level basement extending 80 feet below grade. The highest portion of the structure is 64 feet above grade. The basement houses the radioactive waste treatment facilities, heat-exchangers. pumps, and miscellaneous items. The remainder of the building contai.ns other auxiliary nuclear equipment and associated facilities, hot machine shop, cask handling crane, heating, ventilating, and air conditioning facilities. Refer to figure 3.8-32 thru 3.8-34, inclusive, for the general layout and geometric description of the building. The cask handling crane complies with OSHA Subparagraph N - Materials Handling and Storage of 29 CFR 1910, section 1910.179. 3.8.4.1.3 control Building The control building is a multistory reinforced concrete structure located between the two containment structures with the turbine building at the north side and a separation between it and the fuel handling building which is located south of the control building. Two stories are above grade level and two stories are below grade. The control equipment, plant laboratories, decontamination facilities and locker room are on the grade level floor. The upper floor has office space and air conditioning equipment. The remainder of the building contains the control room, switchgear, battery rooms, communications, computer and cable spreading rooms. Refer to figures 3.8-35 thru 3.8-37, inclusive, for the general layout and building size. r c D D cs es
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VNP 3.8.4.1.4 Fuel Handling Building The fuel handling building is a reinforced concrete structure located at the center of the category I structures comprising the nuclear block for Units 1 and 2 of the Vogtle Nuclear Plant. The fuel building is physically separated from the surrounding structures and has an independent foundation mat. The building contains the new fuel storage area and two spent fuel pools, one for each reactor. The spent fuel pools have thick concrete walls and floor lined on the inside surfaces with stainless steel pla:es for leaktightness. The building superstructure consists of concrete walls and roof system. The new and spent fuel bundles are stored in stainless steel racks spaced 21 inches center-to-center each way. The spent fuel pools are filled with borated water. Each pool is sized to hold fuel bundles for 1-2/3 reactor cores. The fuel handling building has an overhead crane capable of handling such '9avy loads as the fuel cask. Travel of this crane is prevet ed by design from n.oving over the spent fuel pools. Interlocks are provided to prevent the crane from moving over the new fuel area during cask handling operations. A fuel handling crane that runs on rails mounted on the operat-ing floor is provided to handle the new and spent fuel assem-blies. Refer to figures 3.8-39 and 3.8-40 for a plan and sectional view of the building. 3.8.4.1.5 Nuclear Service Cooling Towers The Vogtle Nuclear Power Plant has two nuclear service cooling towers per unit with a storage capacity for 3,640,000 gallons of water in each tower. The cooling towers are partially buried in the soil with a base elevation of 132 feet above datum. The towers are 77 feet apart at their closest point. The water depth in the tower reservoir is 80 feet. The cooling tower is a concrete cylindrical shell with to e of base slab at elevation 132.0 ft and the top of roof at elevation 750.0 ft. It has a total height of 118 ft and an outside diameter of 94 ft. The thickness of the wall is 3 ft, while that of the roof and base slab is 2 ft and 4 ft respectively. There are four fan cylinders extending 14 ft above the roof. At elevation 220 ft there are openings around the wall. The size of the openings is 12 ft by 10 ft with a separation of 2 ft. The fill supports are at elevation 227 ft, while those of cooling fans are at elevation 250 ft. Two D D oo 010 nA1 ULD U,I d b, A S6 3.8-84 y{D 5l e~ ~ .}L m POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP Rg Pipe reactions during normal operating or shutdown conditions, based on the most critical transient or steady state condition. B. Severe Environmental Loads O Severe environmental loads are those loads to be infrequently encountered during the plant life. Included in this category are: E Loads generated by the operating base earthquake (0.6 SSE). W Loads generated by the design wind specified for the plant.
^
. Extreme Environmental Loads Extreme environmental loads are those loads which are credible but are highly improbable. They include:
E' Loads generated by the safe shutdown earthquake and Wt Lo ds generated by the design tornado specified for the plant. They include loads due to the tornado wind pressure, due to tornado-created differential pressures, and due to tornado-generated missiles. D. Abnormal Loads Abnormal loads are those loads generated by a postulated accident within a building and/or compartment thereof. Postulated accidents primarily include high energy pipe ruptures. Included in this category are the following: lh P Pressure equiva]ent static load within or across a compartment and/or building, generated by the postulated accident, and including an appropriate dynamic load factor applied to the peak of the pressure-time curve. T Thermal effects and loads generated by +he a postulated accident. rn N O D U; es en o r qq
- m. dL ,Lr3 a 028 042 S6 3.8-91 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP R Pipe reactions under thermal conditions generated O by the postulated accident. Y r Reaction equivalent static load on the rupture high energy pipe during the postulated accident, g, , and including an appropriate dynamic load factor W applied to the peak of the reaction-time curve. Y. Jet impingement equivalent static load on a 3 structure generated by a rupture of any high energy pipe during the postulated accident, and including an appropriate dynamic load factor applied to the peak of the jet-time load. Y, Missile impact equivalent static load on a structure generated by or during the postulated accident, and including an appropriate dynamic load factor applied to the peak of the missile impact-time curve. 3.8.4.3.8.2 Loading Combinations for Concrete and Steel Structures. For design load combinaticns for structures under this paragraph, refer to tables 3.8-7 and 3.8-8. 3.8.4.4 Design and Analysis Procedures The Category I structures are designed to maintain elastic behavior when subjected to various loading combinations. Plastic design is not intended to be used for any Category I structures analysis. 3.8.4.4.1 Deleted S6 This space has bee:.: left blank intentionally 9 0
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828 043 S6 3.8-92 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21 1979
VNF This space has been left blank intentionally. lS6 0 3.8.4.4.2 Auxiliary Building The static analysis of the auxiliary building is parformed using the theory of elastic frames. The equivalent frame method outlined by the ACI-318-71 code is used. If in certain D**D oW - C "g a g~}I- _J]. _a
. 823 044 S6 3.8-95 POST-CONSTRUCTION PEMilT SUPPLEMENTARY INFOM1ATION - AUGUST 21 1979
VNP areas the modeling of the slabs and walls by the equivalent frame method becomes inaccurate, the finite element will be used in these local areas. The computer program for the analysis is ICES-STRUDL (Structural Design Language), which is outlined in appendix 3F. The analysis of the base slab is performed using plates on elastic foundation technique. Finite element representation of slabs is used. STRUDL uses the stiffness method for the analysis. All different loads are handled by STRUDL, including thermal and tornado loads. The thermal loads are applied as a rise in temperature or a gradient. The tornado loads are applied in three different ways. First a 300 mph applied as windward, leeward, and upward pressure. Second a 3 psi differential prersure between the ins.ide and outside of the building for the auilding portions above ground. Third a missile impingement load as described in subsection 3.3.2. The dynamic analysis of the auxiliary building is accomplished as follows: A dynamic analysis is performed for the purpose of generating seismic rasponse spectrum at different elevations. The mathematical model is a lumped-mass free-free stick model. The modeling technique is described in section 3.7. Soil-structure interaction due to the embedment of the structure is discussed in section 3.7. The time dependent boundary condition described in paragraph 3.7.2.4 is applied at the appropriate nodes of the model to represent the interaction. This boundry condition is a function of accelerations. The output of the dynamic analysis is the time history at different elevations of the structure. Once the time history is obtained, the response spectrum is developed for that elevation. The SPECTRA program, as described in appendix 3F, is used to generate the response spectrum curves which are required for the design of the equipment on the various floors. The maximum inertial forces of the dynamic analysis are applied at the different floor elevations using a three dimensional model of the structure. The stresses in the structural members are then calculated for these forces. 3.8.4.4.3 Control Building The static and dynamic analysis of the control building is the W same as for the auxiliary building. i a 91 0 qe n g n '{ I 9 O ;u "; Qh y b ~ l S6 3.8-96 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP compound. A thick membrane is placed on the soil structure upon which the basin mat is poured. The waterproofing material is used to prevent any intrusion of ground water through the basin shell and mat. The interior surfaces are also waterproofed using a thermoplastic membrane cast in the concrete. The cooling tower and basin are embedded to a considerable depth. This embedment gives rise to additional resistance to overturning from lateral soil pressure. The soil structure interaction is accomplished by the LUSH program. 3.8.5.1.10 Equipment Building The equipment building is supported on the adjacent buildings S6 except for the equipment hatch area which is supported at grade level by means of a conventional slab and grade beau founda-tion system or a mat foundation. 3.8.1.1.11 Pipe and Electrical Cable Tunnels and Electrical Duct Banks The design for the seismic Category I pipe and electrical cable tunnels and electrical duct banks is presently in development stage as stated in paragraph 3.8.4.4.8. For preliminary layout of the underground systems, refer to figures 3.5-1 and 3.5-2'. The discussion as pertains to buried structures in this paragraph will be covered in the FSAR. 3.8.5.2 Applicable Codes, Standards and Specifications Refer to paragraph 3.8.1.2 for the containment and to paragraph 3.8.3.2 for other Category I structures. 3.8.5.3 Loads and Loading Conditions Containment foundation loads and loading combinations are discussed in paragraphs 3.8.1.3 and 3.8.3.3. Foundation loads and loading combinations for other Category I structures are discussed in paragraph 3.8.4.3. 3.8.5.4 Design and Analysi s Procedures Design and analysis procedurei for the containment including the base slab, are discussed in appendix 3R. The basic techniques for analyzing the foundations of Category I structures are by the conventional methods, involving c3 C3 0 d 8 2 8 O g< < g(v es7 - w . - O S6 3.8-123 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP simplifying assumptions, such as are found in the Theory of concrete structures practice. Stresses resulting from local moments, torques, and concentrated reactions, and uniform loading are computed by these methods. These methods are further discussed in paragraph 3.8.3.4. 3.8.5.5 Structural Acceptance Criteria The foundations of all Category I buildings are designed to meet the same structural acceptance criteria as the buildings themselves. These criteria are discussed in paragraphs 3.8.1.5, 3.8.3.5 and 3.8.4.5. The limiting canditions for the foundation medium together with a comparison actual capacity and estimated structure loads are found in paragraphs 2.5.4.10 and 2.5.4.11, 3.8.5.6 Materials Quality Control and Special Construction Techniques The foundations and concrete supports are constructed of concrete using proven methods common to heavy industrial construction. For further discussion, refer to paragraph 3.8.3.6. 3.8.5.7 Testing and Inservice Surveillance Requirements Testing and inservice surveillance are not required and are not planned for foundations of structures or for concrete supports. A discussion of the test program which serves as the basis for the Soils Investigation and Foundation Report in Chapter 2, appendix 2A. e og%" g qS' e na w S6 3.8-124 POET-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
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VNP TABLE OF CONTENTS CHAPTER 6 Section Title Page 6 ENGINEERED SAFETY FEATURES 6.1-1 6.1 GENERAL 6.1-1 6.1.1 SAFETY FEATURES SYSTEMS S6 6.1-1 lS6 6.1.1 SAFETY FEATURES SYSTEMS 6.1-1 6.1.2 OPERATIONAL PELIABILITY 6.1-2 6.2 CONTAINMENT SYSTEMS 6.2-1 O 6.2.1 CONTAINMENT FUNCTIONAL DESIGN 6.2-1 CJ Cs 6.2.1.1 Design Bases en ' cg 7 6.2-1 D f 6.2.1.2 System Design c _, . G.2-5 6.2.1.3 Design Evaluation 6.2-5 6.2.1.4 Testing and Inspection 6.2-29 6.2.2 CONTAINMENT HEAT REMOVAL SYSTEMS 6.2-33 6.2.2.1 Design Bases 6.2-33 6.2.2.2 System Design 6.2-33 6.2.2.3 Design Evaluation 6.2-49 6.2.2.4 Tests and Inspections 6.2-50 6.2.2.5 Instrumentation Application 6.2-53 6.2.3 CONTAINMENT AIR PURIFICATION AND CLEANUP SYSTEMS 6.2-53 6.2.3.1 Design Bases 6.2-54 6.2.3.2 Systems Design 6.2-55 6.2.3.3 Design Evaluation e.2-66 6.2.3.4 Tests and Inspections 6.2-69 823 061 S6 6-i POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP TABLE OF CONTENTS (Continued) Section Title Page 6.2.3.5 Instrumentation Application 6.2-74 6.2.4 CONTAINMENT ISOLATION SYSTEMS 6.2-76 6.2.4.1 Design Bases 6.2-76 6.2.4.2 Systems Design 6.2-81 6.2.4.3 Design Evalaation 6.2-91 6.2.4.4 Tests and Inspections 6.2-91 6.2.5 COMBUSTIBLE GAS CONTROL IN CONTAINMENT 6.2-91 6.2.5.1 Design Bases 6.2-92 6.2.5.2 System Design 6.2-93 6.2.5.3 Design Evaluation 6.2-96 6.2.5.4 Tests and Inspections 6.2-99 6.2.5.5 Instrumentation Application 6.2-99 6.3 EMERGENCY CORE COOLING SYSTEM 6.3-1 6.4 NUCLEAR SERVICE WATER AND COMPONENT COOLING WATER SYSTEMS 6.4-1 6.4.1 NUCLEAR SERVICE COOLING WATER SYSTEM 6.4-2 6.4.1.1 Design Basis 6.4-2 6.4.1.2 System Description 6.4-3 6.4.1.3 Safety Evaluation 6.4-7 6.4.1.4 Tests and Inspections 6.4-7 6.4.].5 Instrumentation Application 6.4-9 6.4.2 COMPONENT COOLING WATER SYSTEM 6.4-9 6.4.2.1 Design Bases 6.4-9 D D oo N O O f S6 6-ii a ~ h] i_d.ia
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N RMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP TABLE OF CONTENTS (Continued) Section Title Page 6.4.2.2 System Description 6.4-13 6.4.2.3 Safety Evaluation 6.4-14 6.4.2.4 Tests and Inspections 6.4-15 6.4.2.5 Instrumentation Application 6.4-15 6.4.3 ULTIMATE HEAT SINK 6.4-15 6.4.3.1 Functions 6.4-15 6.4.3.2 Design Bases 6.4-16 6.4.3.3 System Description 6.4-25 6.4.3.4 Safety Evaluation 6.4-27 6.4.3.5 Tests and Inspections 6.4-29 6.4.3.6 Instrumentation Application 6.4-29 6.4.4 SYSTEM EVALUATION 6.4-30 6.4.4.1 Evaluation Bases 6.4-31 6.4.4.2 Method of Evaluation 6.4-32 6.4.4.3 Results of Evaluation 6.4-33 6.4.4.4 References 6.4-34 6.4.5 FAILURE MODE ANALYSIS 6.4-34 6.4.5.1 Single Failure Criteria 6.4-34 6.5 PENETRATION ROOM FILTRATION SYSTEMS 6.5-1 6.5.2 SYSTEM DESIGN 6.5-la 6.5.2.1 System Description 6.5-la 6.5.2.2 Components "3 D ih 6.5-2
<u ev JL op s 3 o . .
S6 6-iii POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP TABLE OF CONTENTS (Continued) Section Title Page 6.5.3 6.5.4 SYSTEM DESIGN EVALUATION TESTS AND INSPECTIONS 6.5-5 6.5-9 e 6.5.4.1 Component Qualification Tests 6.5-10 6.5.4.2 Component Acceptance Tests 6.5-10 6.5.4.3 Systems Acceptance Tests 6.5-11 6.5.4.4 Periodic Tests Following Acceptance 6.5-13 6.5.5 INSTRUMENTATION APPLICATION 6.5-13 6.
5.6 REFERENCES
6.5-14 S6l 6.6 DELETED S6 G.6-1 6.6 ENCLOSURE BUILDING FILTRATION AND VENT SYSTEM (EBFVS) 6.6-1 6.6.1 DESIGN BASES 6.6-1 6.6.2 SYSTEM DESIGN 6.6-la 6.6.2.1 System Description 6.6-la 6.6.2.2 Components 6.6-3 6.6.3 DESIGN EVALUATION 6.6-6 6.6.4 TESTS 6.6-8 6.6.5 INSTRUMENTATION APPLICATION 6.6-12 6.
6.6 REFERENCES
6.6-13 D MMO e 023 064 S6 6-iv POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
WP mo D TABLE OF CONTENTS (Continued) (3 es
~
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LIST OF TABLES gh, _, k L _.] Table Title Page 6.1-1 Atmosphere Cleanup System Air Filtration and Absorption Units - Compliance with S6 Regulatory Guide 1.5 S6 6.1-2a 6.1-1 Atmosphere Cleanup System Air Filtration and Absorption Units - Compliance with Regulatory Guide 1.5 6.1-2a/b 6.2-1 Energy and Blowdown Time 6.2-3a 6.2-2 Containment Data 6.2-Sa 6.2-3 Mass and Energy Flow into Containment for a Double-Ended Cold Leg Break 6.2-Si 6.. sa Mass and Energy Flow into Containment for a Double-Ended Hot Leg Break 6.2- 3j - 6.2-3b Containment Pressure Capability (Double-Ended Hot Leg Break) 6.2-7a 6.2-3c Blowdown Margin 6.2-7b 6.2-4 Initial Conditions Values 6.2-8a 6.2-5 Containment Safety Features 6.2-8b 6.2-5a Summary of Heat Transfer Correlations used to Calculate Steam Generator Heat Flow in the Satan Code 6.2-9b 6.2-5b Sensitivity of Core Stored Energy to Power Level No. Nodes in the Pellet and Densification 6.2-10a 6.2-6 Structural Heat Sinks 6.2-13a 6.2-7 Material Data 6.2-13b 6.2-8a Containment Margin Design Basis Cases 6.2-14a 6.2-8b Containment Margin Margin Cases 6.2-14b 6.2-9 Double Ended Pump Suction LOCA 6.2-25a 6.2-9a Energy Sources 6.2-25b 82B 065 S6 6-v POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21 1979
VNP TABLE OF CONTENTS (Continued) LIST OF TABLES Table Title Page 6.2-9b Blowdown Energy Balance (106 Btu) 6.2-25c 6.2-9c Reflood Energy Release (106 Btu) 6.2-25e 6.2-10 Energy Distribution for Double Ended Pump Suction Break 6.2-25g 6.2-10A Past-Reflood Mass and Energy Release 6.2-22c 6.2-10B Structural Heat Sinks 6.2-22f 6.2-11 Single Failure Analysis - Containment Spray and Fan Cooler Systems 6.2-34 6.2-12 Containment Heat Removal Systems 6.2-38 6.2-13 Component Design Parameters 6.2-40 6.2-14 Containment Spray System - Code Requirements 6.2-43 6.2-15 Performance Parameters of the Contain-ment and Enclosure Building Air Purifi-cation and Cleanup Filtration System 6.2-56 6.2-16 Filter Design Parameters 6.2-59 6.2-17 Spray Additive System - Design Parameters 6.2-63 6.2-18 Spray Additive System - Code Requirements 6.2-64 6.2-19 Single Failure Analysis 6.2-68 6.2-20 Containment Isolation Valve Information 6.2-93/84 6.2-21 Electric Hydrogen Recombiner Typical Design Parameters 6.2-95 6.2-22 Post Accident Purge System Design Parameters 6.2-97 6.4-1 Safe Shutdown and Cooldown Process Heat Load Requirements i Qp' g 6.4-4 D 6o' gg 1 73 828 Ubb e AL g a S6 6-vi
VNP O TABLE OF CONTENTS (Continued) CJ C's c] c3< 0 LIST OF FIGURES 6-us L) 3 _ 3 Figure Title 6.4-6 Post-LOCA Nuclear Service Cooling Water System Temperature Transient 6.4-7 Time From Shutdown - Days 6.5-1 Mechanical Penetration Room Filtration and Exhaust System 6.5-2 Electrical Penetration Room Filtration and Exhaust System 6.5-3 General Arrangement Penetration Boundary Plan - Elev. 200'-0" 6.5-4 General Arrangement Penetration Boundary Plan - Elev. 180'-0" 6.5-5 General Arrangement Penetration Boundary Plan - Elev. 160'-0" & Elev. 140'-0" 6.6-1 Deleted [S6 6.6-1 Enclosure Building HVAC System 6.6-2 Deleted lS6 6.6-2 Enclosure Building Air Distribution (Sheets 1 Thru 4) 0?3 057 S6 6-xi POST-CONSTRUCTION PERMIT P
VNP CHAPTER 6 ENGINEERED SAFETY FEATURES 6.1 GENERAL Safety teatures are designed to minimize the severity and to mitigate the consequences cf Condition IV Limiting Faults (PSAR Section 15.4) by fulfilling the following safety functions under accident conditions:
- 1. Protect the fuel claddir.g
- 2. Ensure containment integrity
- 3. Minimize containment leakage
- 4. Remove fission products from the containment atmosphere.
- 5. Provide habitable atmosphere for operating personnel.
- 6. Limit radiaactivity discharged to the atmosphere (and f thence the environment) 6.1.1 SAFETY FEATURES SYSTEMS The safety features systems provided tc satisfy the functions listed above are as follows:
Containment isolation system (subsection 6.2.4) Containment spray system (subsection 6.2.2) Containment fan cooler system (subsection 6.2. 2) Containment air purification and cleanup system (subsection 6.2.3) Emergency core cooling system (subsection 6.3) Residual heat removal system (subsection 5.5) Penetration room fiitration system (aubsection 6.5) Combustible gas control in containment [S6 (subsection 6.2.3) Engineered Safety Features Electrical Equipment Rooms HVAC (subsection 9.4.6) Fuel Handling Building post accident filter exhaust ~~ stem (subsection 9.4.5) Control nuom HVAC (subsection 9.4.lA) i <N < 1 0 D c, es g c c} l 023 068 ,a s c6] i _a S6 6.1-1 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP The first safety function is satisfied by the timely, continuous, and adequate supply of borated water to the reactor coolant system and, ultimately, the reactor core. This supply of water is provided by the emergency core cooling system. These systems provide high head (safety injection and centrifugal charging pumps), low head (residual heat removal pumps), injection and accumulator injection immediately following an incident, and low head /high head recirculation in the long term recovery period. The second and third safety C> D< D e, L, 39 cc ,( - u 10 . <0 - ]L L .a 823 069 S6 6.1-la - 'T- ON TRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21 1979
VNP functions are satisfied by the provision of means for con-densing the steam inside the containment, depressurizing the containment following an incident, and maintaining the con-tainment at near atmospheric conditions for an extended period of time. The containment isolation system, spray system, fan S6l cooler system, penetration room vent and filtration system and the electric hydrogen recombiners provide the mears for satisfying these requirements. The fourth safety function is satisfied by providing chemical additives and filters within the containment. These enhance the removal of radioactive iodine from the containment atmos-phere following an incident. The containment air purification and cleanup systems are provided to meet this function. The fifth and sixth safety features are satisfied by the various safety features filtration systems which conforms to Regulatory Guide 1.52 (see table 6.1-1). The safety features systems are designed with sufficient re-dundancy to meet the general design criteria as discussed in section 3.1, 3.2 and appendix 6A. Electrical power for all safety features systems is provided both from offsite sources and from emergency onsite sources as described in sections 8.2 and 8.3, respectively. Safety features are soprated into two independent trains of equal capability. Either train can handle the entire emergency coolant injection and emergency cooling loads; either train can provide the entire containment isolation, containment cleanup, and containment leakage minimization functions. Each train has an independent onsite and offsite power source. Failure of either train cannot affect the other. Both trains are shown in figure 6.1-1, Safety Features Systems. Some of the high and low pressure emergency injection systems use equipment that serves normal functions during normal plant operation or shutdown. Observation of their normal functioning provides monitoring of equipment availability and condition. In cases where equipment is used for emergencies only, systems are designed to permit periodic inspection and tests. 6.1.2 OPERATIONAL RELIABILITY Operational reliability is achieved by using proven components and by conducting tests. All safety features systems are quality items meeting the requirements of 10 CFR 50 appendix B and seismically designed as discussed in chapter 3. Those I D @ 66 - 30 Ql T A fjh S6 6.h U POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
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VNP Table 6.1-1 ATMOSPIIERE CLEANUP SYSTEM AIR FILTRATION AND ADSORPTION . UNITS - COMPLIANCE WITH
'l REGULATORY GUIDE 1.52 ,
~
i. REGULATORY POSITION 4k 41 4m 5a Sb Sc 6a 6b 6c i 4e 4f 49 4h 4i 4j ' S6 . X X X X X X X X X X X X X X
, a X ..i e
X X X X X X X X X X X X
'X X X X X X X X X X X X X X X X X X
k x X x x x x x x x x X x x X X a P s..
^-
b 3 - v dN r i-t a
'~ ' {1~:)
07i l...
S6 6.1-2a/b ..
iv; . 1979 _._ POST-CONSTBUCTION PERMIT SUPPLEMENTitRY INFORMATION - AUGUST 21,
--- ..a y-- _- -
k 21 2m 3a 3b 3c 3d 3e 3f 3g 3h 3i 3j 3k 31 3m 4a 4h 4c 4d X NA X X X X X X X X X X X X X X X X X X NA X X X X X X X X X X X X X X X X X X NA X X X X X X X X X X X X X X X X X X NA X X X X X X X X X X X X X X X X X gggg oS
UNIT DESCRIPTION REGULATORY POSITION 1a 1b 1c 1d le 2a 2b 2c 2d 2e 2f 2g 2h 2i 2j Engineered Safety X X X NA X X X X NA X X X X X X Features Electri-cal Equipment Room HVAC System Fuel liandling X X X NA X X X X NA X X X X X X Building Post Accident Filter Exhaust System hiping and Elec- X X X NA NA X X X NA X X X X X X trical Penetra-tion Room Filter Exhaust System Control Room HVAC X X X NA NA X X X NA X X X X X X System l ME ' X = CompliancP N/A = Not Applicable c,10 O_Q i '3 ua ~-
9 VNP containment structures and the environment was considered to be adiabatic. 6.2.1.2 System Design The analytical technique and design methods used to assure the integrity of the containment internal structure and subcompartments from the effects of LOCA are discussed in paragraph 3.8.1.4 and 3.8.3.4. 6.2.1.3 Design Evaluation Table 6.2-2 provides detailed data on which the containment design is evaluated. Mass addition and energy flow into containment for the double ended pump suction break, the worst break, is given in table 6 2-2 sections IV and V
. . Mass addition and energy flow for the double ended cold leg and the double ended hot leg breaks are given in table 6.2-3 and 6.2-3a respectively.
6.2.1.3.1 Following an accident, the containment air may become radioactively contaminated. The isolation of the containment following an accident therefore is mandatory. . During normal operation, air flowing intermittently from inside the containment to the outer atmosphere is filtered and monitored by the normal containment purge system. Following an accident, the normal purge system is removed from service. Containment leaktightness is accomplished by the redundant quick-closing purge isolation valves located at both supply and exhaust ducts to the containment. Additionally, the piping and electrical penetration filter exhaust systems maintain a S6 subatmospheric pressure in the penetration rocms following a loss-of-coolant accident. These systems are discussed further under subsection 6.2.3 and section 6.5 respectively. This space has been left blank intentionally 0 0 Cu cu c3 b D 1[ f w . - 1 02B 074 36 6.2-5 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGU
VNP Table 6.2-3a (Continued) MASS AND ENERGY FLOW INTO CONTAINMENT FOR A DOUBLE-ENDED HOT LEG BREAK Time Mass Rate Energy Rate (sec) (lbs/sec) (Etu/sec) 19.8 0. O. 20.9 0. O. 22.18 1231.1 99.83 E+04 27.8 4365.5 158.70 E+04
- 50. 1928.3 105.82 E+04
*62.800 900.8 84.93 E+04 62.801 194.0 23.09 E+04 100. 176.0 20.93 E+04 200. 143.56 17.08 E+04 500. 104.67 12.46 E+04 1000. 79.20 9.43 E+04 2000. 60.60 7.20 E+J4 5000. 42.05 5.36 E+04 10000. 35.78 4.26 E+04
- Time that quench front reaches 8 ft elevation in the core.
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VNP S6 6.2.1.3.2 The piping and electrical penetration filter exhaust systems are designed to operate during emergency situations. The systems are designed as seismic Category I. All equipment is provided with 100 percent redundancy, including electrical buses. Each train in these systems is connected to one emergency bus which assures operation follow-ing loss of offsite power. 6.2.1.3.3 Blowdown Containment Pressure Analysis 6.2.1.3.3.1 Results. The results of the pressure transient analysis of the containment for the blowdown phase of the loss-of-coolant accident are shown in figure 6.2-1 sheet 1 and 2. The cases examined in thi- analysis determine the effects of the full range of large actor coolant break sizes up to and including a double en( rupture. Cases illustrating the sensitivity to break location are also shown. All of these cases show that the containment pressure will remain below design pressure with margin. This blowdown margin can be represented in terms of energy or pressure. Figure 6.2-2, the plot of containment pressure versus internal energy of the containment atmosphere, can help quantize the available margin. Point A on this figure represents the blowdown pressure; point B corresponds to the containment design pressure. The increase in energy necessary to go from point A to point B in this figure represents the energy margin available in the containment design. Since energy transferred to the containment from the core is in the form of steam, the total transferred core energy corresponding to allowed energy addition can be calculated as follows: h O core
- C allowed D k O core = Core energy release a
g4 0' 3 I h fg
= Latent beat of vaporization w - -
h = Steam enthalpy g Q allowed = Permissible increase in containment energy This allowable value of energy which could be transferred from the core to the containment without increasing the transient containment pressure to design pressure can be compared to the 000 S6 6.2-6 POST-CONSTRULTION PERMIT CCPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP 6.6 DELETED lS6 1 Q9 D D W o Ju ar , 1
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0 02B 078 S6 6.6-6 q , T N - AUGUST 21 1979
This figure is deleted by Supplement No. 6 D
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U .. i ) U! i Figure 6.6-1 POST-CONSTRUCTION PEIU4IT SUPPLEMENTARY INFORMATION - AUGUST 21 1979
4"% This figure is d'eleted by Supplement No. 6
"[1L -
o 9 m3 o . d) k _ 3 023 000 Figure 6.6-2 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21 1979
VNP D D CJ c9 JU
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TABLE OF CONTENTS D lh l[ a D3. L n CHAPTER 7 Section Title Page 7 INSTRUMENTATION AND CONTROLS 7.1-1 INTRODUCTION 7.1-1 ) 7.1 7.2 REACTOR TRIP SYSTEM 7.2-1 7.3 ENGINEERED SAFETY FEATURES ACTUATION S6 SYSTEM S6 7.3-1 7.3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM . 3-1 7.3.4 MANUALLY-ACTUATED ENGINEERED SAFETY FEATURES SYSTEM EQUIPMENT i 7.3-4 7.3.4,1 Electric dydrogen Recombiners 7.3-4 7.3.4.2 Combustible Gas Control in Containment 7.3-5 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 7.4-1 7.4.2 ANALYSIS 7.4-1 7.5 SAFETY RELATED DISPLAY INSTRUMENTATION 7.5-1 7.5.2 ANALYSES 7.5-1 7.5.3 SAFETY FEATURES EQUIPMENT BYPASS INDICATION 7.5-1 7.6 ALL OTHER SYSTEMS REQUIRED FOR SAFETY 7.6-1 7.7 PLANT CONTROL SYSTEMS 7.7-1 LIST OF TABLES Tables for this chapter are presented in RESAR-3, chapter 7. S':S 03i S6 7-i
VNP LIST OF FIGURES Figures for this chapter are presented in RESAR-3, chapter 7. O O
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O O 02B 032 S6 7-ii POST-COMSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP 7.3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM This section is presented in RESAR-3, Section 7.3 with the following modifications and additions. 7.3.1.1.1.2
- a. Per RESAR-3
- b. Per RESAR-3
- c. Containment air recirculation fans and filtration system which serve to cool the containment and limit the potential for release of fission prodt. cts from the containment by reducing the pressure following an accident.
- d. Nuclear service water pump which provides cooling water to the component cooling system heat exchangers and is thus the heat sink for containment cooling.
- e. A :iliary feedwater pumps.
- f. Containment dome circulator fans which will prevent stratification of gases in the containe.ent dome and the possible accumulation of a dangerous mixture of hydrogen and air from occurring.
9 Containment cooling units which serve to reduce pressure by cooling the containment following an accident.
- h. ?enetration room filtration system which will limit S6 release to the environment of radioisotopes that have leaked from the containment into the penetration rooms, hh 1 D ou,si
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U.0 UUJ S6 7.3-1 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21 1979
VNP 7.3.1.1.2 Analog Circuitry Numbers 1 through 4 are per RESAR-3.
- 5. ~ Containment air coolers The containment air coolers cooling water discharge flow is alarmed in the control room if the flow is low or if the differential (inlet versus outlet) flow is high or low. Cooling water flow through each cooler is recorded in the control room. The transmitters are outside the reactor containment. Local pressure and temperature indicators are provided outside the containment to monitor the cooler cooling water discharge flov. A1: cooler air flow it monitored
, ~
by an alarm tunction in the control room which is
-actuated P low discharge. air flow. In addition a comnon .adiation monitor exists for exit cooling water ficw wbAch will actuate a control room alarm in the event of high radiation. The faulty ccoler can be detected locally by manually valving each cooler out in turn.
The instrument lines which sense cooling water flow do not penetrate the containment but are tapped into the p):ocese cooling water lines at orifice plates located outside of the autmaatic isolation valves which are external to the containment. These valves are installed lu
. accordance with 10CFR50, Criterion 57. The low discharge air flow signal is provided by an electric switch contact.
) t) . (Per RESAR-3)
(
- 7. Deleted O S6 This space has taen left blank intenti a 9 2 e
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,- , -; E6 7.3-2 POST-CONSTRDCTION PEPMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP TABLE OF CONTENTS C11 APTER 8 Section Title Page 8 ELECTRIC POWER 8.1-1 E.1 INTRODUCTION 8.1-1 8.1.1 UTILITY GRID DESCRIPTION 8.1-1 8.1.2 ONSITE POWER SYSTEM DESCRIPTION 8.1-1 8.1.3 SAFETY FEATURES 8.1-2 8.1.4 DESIGN CRITERIA 8.1-2 8.2 OFFSITE POWER SYSTEM 8.2-1 8.2.1 SYSTEM DESCRIPTION 8.2-1 8.2.2 ANALYSIS 8.2-2 8.2.2.1 Electrical Power Systems 8.2-2 8.2.2.2 Inspection and Testing of Electrical Power System 8.2-3 8.2.2.3 Analysis of Loss Of Offsitt _ower 8.2-3 8.3 ONSITE POWER SYSTEM 8.3-1 8.3.1 AC POWER SYSTEM 8.3-1 8.3.1.1 Description 8.3-1 8.3.1.2 ._nalysis 8.3-7 8.3.2 DC DISTRIBUTION SYSTEMS 8.3-23 8.3.2.1 Description gn c3 8.3-23 g 8.3.2.2 Analysis ev eg f 8.3-24
* ~
0 ~%~T O - _ _)- O UL"; b' 00~ ua3 S6 8-i POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP TABLE OF CONTENTS (Continued) LIST OF TABLES Table Title Page Safety Loads and Functions S6 8.1-3 S6l 8.1-1 8.1-1 Safety Loads and Functions 8.1-3 8.3-1 Emergency Electrical Loading Require-S6 ments for Loss of Coolant Accident and Loss of Offsite Power S6 8.3-3 8.3-1 Emergency Electrical Loading Require-ments for Loss of Coolant Accident and Loss of Of f site Power 8.3-3 8.3-2 Emergency Electrical Loading Require-S6 ments for Blackout Conditions S6 8.3-5 8.3-2 Emergency Electrical Loading Require-ments for Blackout Conditions 8.3-5 8.3-3 Failure Mode and Effect Analysis for Auxiliary AC Power System 8.3-9 0 8.3-4 Failure Mode and Effect Analysis for Auxiliary DC Power System 8.3-25 8.3-5 120 V AC Vital Power System Loads (Refer to figure 8.3-8 for one line) 8.3-31 8.3-6 Safety Features 125 V DC Loads (Refer to figure 8.3-8 for one line) 8.3-32 O O B D k
@@N h b S6 8-ii "- 4T TO P RM T SUPPLEMENTARY INFORMATION - AUGUST 21 1979
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VNP QwI C3 D D f Table 8.1-1 eg j sh _ )L _; SAFETY LOADS AND FUNCTIONS SAFETY LOADS FUNCTION POWER Component Cooling Provide cooling water to NSSS AC
'; rater Pumps equipment Residual lleat Remove reactor heat during hot AC Removal Pumps shutdown Charging Pumps Provide emergency core cooling AC during emergency shutdown Safety Injection Provide emergency core cooling AC Pumps during emergency shutdown Containment Spray Provide cooling spray in contain- AC Pump ment during LOCA Nuclear Service Provide water for component AC Cooling Water Pumps cooling water system Containment Fans For cooling containment after AC accident Spent Fuel Pool Cool Spent Fuel Pool AC Coolant Pump Hydrogen Recombiner Maintains a safe level of lifdrogen AC in Containment Emergency Air Maintains a safe air operating AC Conditioning temperature Fire Pump To provide water to extinguish AC fires Reactor Service Cool service cooling water AC Cooling Twr. I.D.
Fans Motor Control Centers Provide power for M.O.V., small AC motors, fans, heaters, and small pumps associated with safety related equipment. Penetration Room Minimize containment leakage AC Filtration & Vent n", nO7 S6 U ._ d Ud/ S6 8.1-3 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21 1979
VNP Table 8.1-1 (Continued) SAFETY LOADS AND FUNCTIONS SAFETY LOADS FUNCTION POWER Fuel Handling Building Mitigate the effects of AC Filtration & Vent fuel handling accident Containment H 2 Purge Control Post-LOCA H 2 AC System Concentration Safeguard Logic Initiates safety injection DC Relay System signal Reactor Protection Prevents reactor from DC Logic Relay System operating in unsafe condi-tions Auxiliary Relay Auxiliary relays for process DC Cabinet control and nuclear instru-mentation Reactor Trip Trips reactor DC Switchgear Waste Disposal Control panel for waste DC Panel disposal system Instrument Power Supplies power to the vital DC Supply Inverter instrtunent busses NSSS Solenoid Controls flow of the NSSS DC Valves: (pneumatic valves with CVCS, RCS solenoid actuators) SIS Sampling System Distribution Panel Supplies power to emergency DC lighting and protective relay panel e g1
- P;dI^
un m e S6 8.1-4 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
O D fl Table 8.3-1 t o o _0 ! g O EMERGENCY ELECTRICAL LOADING REQUIREMENTS FOR D A LOSS-OF-COOLANT ACCIDENT AND LOSS OF OFFSITE POWER o . . _] No. HP Each Auto. Delay Time After Installed Nameplate Sequence Safeguard Signal Component Per Unit Rating Start is Actuated Centrifugal charging pumps 2 600 Yes 10 Sec. Safety injection pumps 2 400 Yes 15 Sec. Residual heat removal pumps 2 400 Yes 20 Sec. Containment spray pumps 2 400 Yes 25 Sec. Component cooling water pumps 4 800 Yes 30 Sec. Nuclear service water pumps 4 1100 Yes 35 Sec. " E d Valves (motor operated) - 175 Yes 10 Sec. Nuclear service cooling tower fans 4 125 Yes 10 Sec. Battery chargers 3 45 kVA Yes 10 Sec. I Control Room Outside Air Co Supply 2 10 Yes 15 Sec. i' . 3 w Control Room Air Condi-tioning 2 60 Yes 15 Sec. Control Room Heating Coil 2 120 Yc s 15 Sec. lS6 Penetration Room Recirculation
- Exhaust 2 15 Yes 15 Sec.
m 8 8 A Table 8.3-1 (Continued) o
$ EMERGENCY ELLCTRICAL LOADING REQUIELMENTS FOR j LOSS-OF-COOLANT ACCIDENT AND LOSS OF OFFSITE POWER C
o 8 y No. HP Each Auto. Delay Time After Z Installed Nameplate Sequence Safeguard Signal m Component Per Unit Rating Start is Actuated tn Penetration Room Exhaust 2 1.5 Yes 15 Sec. 8 m Containment Cooler Fan No. 1 2 150 Yes 15 Sec. C Containment Cooler Fan No. 2 2 150 Yes 15 Sec. tm
$m g Ilydrogen Recombiner 2 75 kW Yes 15 Sec.
2
$. Post-Accident Containment g g Purge Exhaust Fan 2 1 Yes 15 Sec. y y D' Post-Accident Containment y Purge Supply Fan 2 0.75 Yes 15 Sec.
Cavity Cooling Supply 2 10 Yes 15 Sec. 8 o Upper Dome Circulator 2 20 Yes 15 Sec. I Upper Dome Circulator 2 20 Yes 15 Sec.
> g(3 i d @ w Refueling Pool Vent Unit 2 3 Yes 15 Sec. ,
iM CD Spent Fuel Pool Post-u
'- O Accident Exhaust 2 15 Yes 15 Sec. '
'g Electrical Equipment Room b d
[ d-Exhaust Air Conditioning 2 15 Yes 15 Sec. b a P
n g VNP
$ b3 '
Table 8.3-1 (Continued) g EMERGENCY ELECTRICAL LOADING REQUIREMENTS FOR LOSS-OF-COOLANT ACCIDENT AND LOSS OF OFFSITE POWER Delay Time No. HP Each Auto. After Safe-Installed Nameplate Sequence guard Signal Component Per Unit Rating Start is Actuated Battery Room 2 1 Yes 15 Sec. Air Condi-tioning Emergency sub- 2 125 Yes 20 Sec. merged deep well make-up Pump Emergency - 40 kW Yes Immediately lighting Reciprocating 1 200 No* - charging pump Boric acid 3 15 No* 60 Min. transfer pumps Pressurizer 2 Banks / 350 kW/ No* - heaters Unit Bank Boron injection 2 6 kW No* - tank heater Heat tracing No* 5 Min. Instrumentation - 5 kW Yes Immediately and control Containment No* Ventilation fans Spent Fuel Pool 1 250 No* - Cooling Pumps Auxiliary Feed- 2 500 Yes 40 Sec. water Pumps O
- Manual start when required. Train A = 7438 hp q7q' Q {j 't Train B = 7438 hp b S6 8.3-4 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP Table 8.3-2 EMERGENCY ELECTRICAL LOADING REQUIREMENTS FOR BLACKOUT CONDITIONS (Sheet 1 of 2) Blackout HP Each Delay No. Name- Auto. Time Installed plate Scquence After Component Per Unit Rating Start Elackout , Centrifugal Charging 2 600 Yes 10 Sec. Fumps Coraponent Cooling 4 800 Yes 30 Sec. Water Pumps Nuclear Service Water 4 1100 Yes 35 Sec. Pumps Valves (motor - 175 Yes 10 Sec. operated) Instrument 1 tion and - 5 kW Yes Immediately Control Nuclear Service Cool- 4 125 Yes 10 Sec. ing Towers Battery Chargers 3 45 kVA Yes 10 Sec. Control Room Outside 2 10 Yes 15 Sec. Air Supply Control Room Air 2 60 Yes 15 Sec. Conditioning Control Room Heating 2 120 Yes 15 Sec. lS6 Penetration Room 2 15 Yes 15 Sec. Recirculation Exhaust D 0) e>cw10 o
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D M WP [ Table 8.3-2 o w w T ra p _]_ { g
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Q EMERGENCY ELECTRICAL LOADING REQUIREMENTS FOR BLACKOUT CONDITIONS (Sheet 2 of 3) Blackout HP Each Delay No. Name- Auto. Time Installed plate Sequence After Component Per Unit Rating Start Blackout Penetration Room 2 1.5 Yes 15 Sec. Exhaust Containment Cooler 2 150 Yes 15 Sec. Fan No. 1 Containment Cooler 2 150 Yes 15 Sec. Fan No. 2 Hydrogen Recombiner 2 75 kW Yes 15 Sec. Post-Accident 2 1 Yes 15 Sec. Containment Purge Exhaust Fan Post-Accident 2 0.75 Yes 15 Sec. Containment Purge Supply Fan Cavity Cooling Supply 2 10 Yes 15 Sec. Upper Dome Circulator 2 20 Yes 15 Sec. Upper Dome Circulator 2 20 Yes 15 Sec. Refueling Pool Vent 2 3 Yes 15 Sec. Unit Spent Fuel Pool Post- 2 15 Yes 15 Sec. Accident Exhaust Electrical Equipment 2 15 Yes 15 Sec. Room Exhaust Air Conditioning Battery Room Air 2 1 Yes 15 Sec. Conditioning O S6 8.3-Sa 828 093 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP TABLE OF CONTENTS (Continued) LIST OF TABLES Table Title Page 9.5-1 Failure Mode and Effects Analysis of Diesel Generator Fuel Oil System 9.5-10 9.5-2 Fire Protection for Safety Related S6 Equipment S6 9.5-11 9.5-2 Fire Protection for Safety Related Equipment 9.5-11 9.5-2A Fire Protection Tabulation Table 9.5-2a 9.5-3 Failure Mode and Effect Analysis for Emergency Lighting System 9.5-13 9.5-4 Failure Modes and Effects Analysis of the Fire Protection System 9.5-16 rN cN 0 D cv Cu
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- S6 9-vii POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP TABLE OF CONTENTS (Continued) LIST OF FIGURES OOI O T D Figure Title w _ _ _.3 9.1-1 Spent Fuel Pit Cooling System 9.2-1 Plant Makeup Water and Treatment System 9.3-1 Compressed Air System 9.3-2 Boron Recycle System Process Flow Diagram 9.3-3 Process Sampling System 9.3-4 H 2 Storage Facility for Units 1 & 2 9.4-1 Control Room Locker, Offices and Lab HVAC System 9,4-2 Electrical Equipment Room HVAC System 9.4-3 Fuel Handling Building HVAC System 9.4-4 P&I Diagram Auxiliary Bldg. (Radwaste) HVAC System 9.4-6 Turbine Building HVAC System No. 1571 & 1572 9.4-7 P&I Diagram Diesel Generator Building Ventilation System 9.5-1 Piping and Instrumentation Diagram for Fire Protection Water System 9.5-2 Diesel Generator Fuel Oil System 9.5-3 Emergency Lighting One Line Failure Mode Analysis 9.5-4 Diesel Generator Cooling Water System 9.5-5 Diesel Generator Air Start System 9.5-6 Diesel Generator Lube Oil System 9.5-7 Fire Protection Piping Layout Plan Elev 200'-0" & Above 9.5-8 Fire Protection General Arrangement Plan S6 9-viii 823 095 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
M O tn a o e Table 9.5-2 O y FIRE PROTECTION FOR SAFETY RELATED EQUIPMENT H M h Accessibility e F4 Fit a Toxic Protection Detection Combustion Safety Related Equipment System Operation Device Heat Radiation Products 1.egend M Control room support area air conditioning ur.it E. F M HD 0 0 0 Fire Protection System M h Post-LOCA purge unit E M HD, SD 0 0 0 C - Std. wet pipe H sprinklers 8 Control room E. F M HD, SD 0 0 0 m D - Fixed pipe CO system lS6 C Normal-emergency containment air cooling units F M HD, SD 0, X 0, X 0 2 M **Ei"~ M Component cooling water heat exchanger C. H AM S 0 0 0 E - Portable CO2 tinguisher m Ch Turbine driven auxiliary feedwater pump C, H A S 0 0 0 F - Portable dry chemical M extinguisher Z Diesel driven auxiliary feedwater pump A HD, SD 0 0 ge D 0
- p. C - CO2 hose reels Z n 4160 V switchgear G H SD 0 0 0 M N
K (3 H - Firewster hose station g Penetration rooma filtration systen EF M HD 0 0 0 g Detection Devices M Cable spreading room D M HD 0 0 P O S - Sprinkler head Accumulator tanks F M HD, SD 0, X 0, X 0 HD - Heat detector 8 Component cooling water pumps C, H A, 11 9 0 0 H SD - Smoke detector h Emergency air conditioning system EF M HD 0 0 0 Operation I MCC and switchgear rooms C M SD 0 0 P p A - Automatic c 125 VDC battery rooms E M SD 0 0 0 (- } O M - Manual @ 'y w Communication room E H HD 0 0 0 a g Ac:essibility Electric and air conditioning equipment room C, E F M HD, SD 0 0 P i i N .-- 0 - No special protective H ( RHR heat exchangers C, H AM S O O O Jevice L b ') H O O' Containment spray heat exchangers C, H A. M S 0 0 0 P - Protective device provided
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VNP TABLE OF CONTENTS CHAPTER 12 Section Title Page 12 RADIATION PROTECTION 12.1-1 12.1 SHIELDINr 12.1-1 12.1.1 DESIGN OBJECTIVES 12.1-1 12.1.2 DESIGN DESCRIPTION 12.1-1 12.1.2.1 Primary Shield 12.1-3 12.1.2.2 Secondary Shield 12.1-4 12.1.2.3 containment Shield 12.1-4 12.1.2.4 Control Room Shield 12.1-4 12.1.2.5 Auxiliary Shield 12.1-5 12.1.2.6 Spent Fuel Shielding 12.1-5 12.1.2.7 Valve Stations 12.1-6 12.1.2.8 Radioactive Piping 12.1-6 12.1.2.9 Penetrations 12.1-7 12.1.2.10 Temporary Shielding 12.1-7 12.1.2.11 Maintenance considerations 12.1-8 12.1.2.12 Materials 12.1-8 12.1.3 SOURCE TERMS 12.1-8 12.1.4 AREA MONITORING 12.1-13 12.1.4.1 System Circuit Description 12.1-16 12.1.4.2 Calibration and Maintenance P99D 12.1-16 66 g* I njan sli 823 097 Str0 dd0u ' S6 12-i POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21 1979
VNP TABLE OF CONTENTS (Continued) Section Title Page 12.1.5 OPERAr2ING PROCEDURES 12.1-17 12.1.5.1 External Exposure 12.1-17 12.1.5.2 Experience 12.1-17 12.1.6 ESTIMATES OF EXPOSURES 12.1-18 12.1.6.1 Exposures in the Plant Buildings and on the Plant Site 12.1-18 12.1.6.2 Exposures at the Site Boundary and in Uncontrolled Areas 12.1-18 12.1.6.3 Comparison with Other Relevant Plants 12.1-19 12.1.6.4 Justification for the Shielding Employed 12.1-19 12.2 VEF"tILATION 12.2-1 12.2.1 DESIGN OBJECTIVES 12.2-1 12.2.2 DESIGN DESCRIPTION 12.2-1 12.2.2.1 Containment Building 12.2-1 12.2.2.2 Penetration Rooms 12.2-1 S6l 12. 2. 2. 3 Deleted S6 12.2-1 12.2.2.3 Enclosure Building 12.2-1 12.2.2.4 Control Room Area 12.2-1 12.2.2.5 Fuel Handling Building 12.2-1 12.2.2.6 Radwaste Area 12.2-1 12.2.2.7 Battery Room D 12.2-2 12.2.2.8 Diesel Generator Building d 12.2-2 12.2.2.9 Turbine Building 5 .2-2 12.2.2.10 Plant Vents 12.2-2 ngo C'.0 oco V/O 9 S6 12-ii POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP. 12.2 VENTILATION 12.2.1 DESIGN OBJECTIVES The plant is designed to provide maximum safety and convenience for operating personnel with equipment arranged in zones so that areas of potential contamination are separated from clean areas. The plant ventilation systems are designed to provide a suitable environment for equipment and personnel. 12.2.2 DESIGN DESCRIPTION 12.2.2.1 Containment Building Ventilation for the containment building is discussed in section 6.2. 12.2.2.2 Penetration Rooms Ventilation for the penetration rooms is discussed in section 6.5. 12.2.2.3 Deleted lS6 12.2.2.4 Control Room Area Ventilation for the control room area is discussed in section 9.4.1. 12.2.2.5 Fuel Handlf ag Building Ventilation for the fuel handling building is discussed in section 9.4.5. 12.2.2.6 Radwaste Area Ventilation for the radwaste area is discussed in section 9.4.3. C3 C3 9 9
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VNP 12.2.2.7 Battery Room Ventilation for the battery room is discussed in section 9.4.6. 12.2.2.8 Diesel Generator Building Ventilation for the diesel generator building is discussed in
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section 9.' 12.2.2.9 Turbine Building Ventilation for the turbine Lailding is discussed in section 9.4.4. 12.2.2.10 Plant Vents Plant vents in seismic Category I structures are designed for seismjc Category I and for tornado conditions. 12.2.3 SOURCE TERMS Design estimates of normal annual liquid leakage volumes for a single unit are presented in section 11.2. Airborne radioactivity will be calculated based on an expected distribution of leakage into the containment and auxiliary building. All noble gases and a portion of the halogens and particulates are assumed to become airborne. 12.2.4 AIRBORNE RADIOACTIVITY MONITORING The ventilation airborne radioactivity monitoring system provides radiation measurements, indications, records, alarms and controls at selected locations to verify compliance with AEC applicable limits to detect and control abnorma occurrences within the plant. Monitors are located in ventilation systems where persani.el exposure to radiation is most likely, gaseous effluent streams, and in the process line of the waste gas processing system. A detailed description of the airborne radioactivity monitors is given 111 paragraph 11.4.t.2.2 and tabulated in table 11.4-3. O o a 48@ 828 100 S6 12.2-2 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP D TABLE OF CONTENTS f# - fd D) b )_ g CHAPTER 15 g _]_ [ g Section Title Page 15 ACCIDENT ANALYSIS 15.1-1 15.1 CONDITION I - NORMAL OPERATION AND OPERATIONAL TRANSIENTS 15.1-1 15.2 CONDITION II - FAULTS OF MODERATE FREQUENCY 15.2-1 15.2.8 LOSS OF NORMAL FEEDWATER 15.2-1 15.2.8.1 Identification of Causes and Accident Description 15.2-1 15.2.8.2 Analysis of Effects and Consequences 15.2-2 15.2.9.4 Environmental Consequences of a Postu-lated Loss of AC Power to the Station Aux 111 aries 15.2-4 15.2.14 REFERENCES 15.2-6 15.3 CONDITION III - INFREQUENT FAULTS 15.3-1 15.3.5.3 Environmental Consequences of a Postu-lated Waste Gas Decay Tank Rupture 15.3-1 15.3.5.4 References 15.3-4 15.4 CONDITION IV - LIMITING FAULTS 15.4-1 15.4.1.3 Environmental Consquences of a Postu- S6 lated Loss of Coolant Accident S6 15.4-2 15.4.1.3 Environmental Consequences of a Postu-lated Loss of Coolant Accident 15.4-2 15.4.2.4 Environmental Consequences of a Postu-lated Steam Line Break 15.4-17 15.4.3.4 Environmental Consequences of a Postu-lated Steam Generator Tube Rupture 15.4-19 15.4.5 FUEL HANDLING ACCIDENT 15.4-22 15.4.5.4 Environmental Consequences 15.4-22 823 '.01 S6 15-i POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21 1979
VNP TABLE OF CONTENTS (Continued) Section Title Page 15.4.6 ENVIRONMENTAL CONSEQUENCES OF A POSTULATED ROD EJECTION ACCIDENT 15.4-24 15.4.6.1 Model 15.4-24 15.4.6.2 Assumptions 15.4-25 15.4.6.3 Ultra Conservative Analysis 15.4-28 15.4.6.4 Result _s_ 15.4-29 15.
4.7 REFERENCES
15.4-29 O
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- RY NFORMATION - AUGUST 21, 1979
VNP TABLE OF CONTENTS (Continued) LIST OF TABLES Table Title ,Page 15.3-1 Waste Gas Decay Tank Inventory 15.3-2 15.4-11 Parameters Used in LOCA Analysis SG 15.4-5 15.4-llA Iodine and Noble Gas Inventory in Reactor Core and Fuel Rod Gaps S6 15.4-6 15.4-llB Noble Gas and Iodine Inventory in S6 the Containment Atmosphere Immediately After LOCA and Available for Leakage S6 15.4-7 15.4-11C Activity Releases to Atmosphere from Loss of Coolant Accident S6 15.4-9 j 15.4-11 Parameters Used in LOCA Analysin 15.4-5 15.4-lla Parameters Used to Determine Hydrogen Purging Activity Release 15.4-13 15.4-]lb Doses From Containment Purging to Control Hydrogen 15.4-14 15.4-12 Potential Offsite Doses Due to Accidents Vogtle Nuclear Plant S6 S6 15.4-10 15.4-12 Loss of Coolant Accident 15.4-10 15.4-12A LOCA Cases Analyzed 15.4-llb 15.4-12B LOCA Off-Site Doses 15.4-lle 15.4-13 Parameters Used in Rod Ejection Accident Analysis 15.4-26 D
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S6 15-iii POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21 1979
VNP TABLE OF CONTENTS (Continued) LIST OF FIGURES Figure Title 15.2-29 Transient Response Following a Loss of Normal Feedwater 15.2-52 Loss of AC Pover - Thyroid Doses 15.2-53 Loss of AC Power - Whole Body Gamma and Beta Doses 15.4-59 Schematic of Leakage Path for 90% Mixing Case 15.4-60 Schematic of Leakage Path for Short Circuit Case 15.4-61 Steam Line Break - Whole Body Gamma Beta Dose 15.4-62 Steam Line Break Accident - Thyroid Doses 15.4-63 Steam Generator Tube Rupture - Whole Body Gamma and Beta Doses 15.4-64 Steam Generator Tube Rupture - Thyroid Doses 15.4-65 Rod Ejection Accident Thyroid Dose - No Core Melting 15.4-66 Rod Ejection Accident Beta Doses - No Core Melting 15.4-67 Rod Ejection Accident Gamma Doses - No Core Melting 15.4-68 Rod Ejectica Accident Thyroid Doses - 0.25% Core Melting 15.4-69 Rod Ejection Accident Beta Dose - 0.25% Core Melting 15.4-70 Rod Ejection Accident Gamma Dose - 0.25% Core Melting S6 15-iv g i @82B 104 ' POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
D lh cv ev JL VNP g . - - O - . _J 15.4 CONDITION IV -- LIMITING FAULTS Unless otherwise indicated, the applicable data for this section are presented in RESAR-3, section 15.4. The information for the four-loop plant is applicable to the VNP. NOTE: The following paragraph should be read in place of the first paragraph on page 15.4-23 of RESAR-3. Figure 15.4-29 shows the required total peaking factor at the license application power rating to meet the AEC Interim Acceptance Criteria for ECCS as a function of calculated peak containment pressure for a double ended cold leg break. The calculated peak containment pressure for the double ended cold leg break is reported in section 6.2.1. Using this pressure in figure 15.4-29, the maximum allowable linear power and total peaking factor at a core power of 3411 MWt for which the ECCS will meet the AEC Interim Acceptance Criteria can be obtained. Note that the peaking factor in figure 15.4-29 is based on ECCS analysis at containment pressure values c f 90 percent of the calculated peak containment pressure for blowdown and 80 percent of the peak for reflood, as specified by the AEC Interim Policy Statement. NOTE: The first paragraph in RESAR-3, page 15.4-42 should read as follows: Core Power and Reactor Coolant System Transient Figure 15.4-37 shows the reactor coolant system transient and core heat flux following a main steam pipe rupture (complete severance of a pipe) outside the containment, downstream of the flow measuring nozzle at initial no load condition (case a). The break assumed is the largest break which can occur anywhere outside the containment either upstream or downstream of the isolation valves. Offsite power is assumed available such that full reactor coolant flow exists. The transient shown assumes a steam release from only one steam generator. Should the core be critical at near zero power when the rupture occurs the initiation of safety injection by high differential pressure between any steam line and the remaining steam lines or by high steam flow signals in coincidence with either low-low reactor coolant system temperature or low steam line presEure will trip the reactor. Steam release from more than one steam generator will be prevented by automatic trip of the bi-directional isolation valves in the steam lines by the high steam flow signals in coincidence with either low reactor coolant system temperature or low steam line pressure. Even with the failure of one valve, release is limited to no more than 5 seconds for C28 105 S6 15.4-1 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP the other steam generators while the one generator blows down. The steam line isolation valves are designed to be fully closed in less than 5 seconds. The following paragraphs on environmental consequences of postulated accidents are added to RESAR-3, section 15.4. 15.4.1.3 Radiological Consequences of a Postulated Loss Of Coolant Accident (See Subsection 6.2.1.3). The results of analyses presented in this section demonstrate that the amounts of radioactivity released to the environment in the event of a loss of coolant accident do not result in doses which exceed the guideline values specified in 10 CFR 100. The analyses performed is based on Regulatory Guide 1.4, Revision 2. The parameters used for this analysis are listed in table 15.4-11. In addition, an evaluntion of the offsite dose resulting from purging the containment for hydrogen con-trol, and an evaluation of the offsite doses resulting from recirculation loop leakage, are presented in subsection 15.4.1.3.9. 15.4.1.3.1 Fission Product Release to the Containment The calculation of potential offsite doses resulting from a loss of coolant accident are based on the conservative fission O S6 product releases recommended by Regulatory Guide 1.4. One hundred percent of the core noble gas inventory and 25 per-cent of the core iodine inventory is assumed to be immediately available for leakage from the primary containment. Ninety one percent of the halogen activity available for release is assumed to be in elemental form, 4 percent in methyl form, and 5 percent in particulate form. The total core noble gas and iodine inventories ara given in table 15.4-llA, while the activity in the containment atmosphere immediately following the LOCA and available for leakage is shown in table 15.4-llB. 15.4.1.3.2 Containment Model The activity released from the containment was calculated with a two-volume model to represent sprayed and unsprayed regions of the containment. g @@ O.20 .06 S6 15.4-2 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP The initial instantaneous release of fissicn products to the containment is assumed to be the only source of activity in the containment. The activity change with respect to time in each of the two well-mixed volumes is described by the following equations: ky da A a - 2ay+ 21 a 2 (15.4-1) i=_ y,3 y dt 1 2 j=1 k2 da 2=- A 2,j"2 ~ 1# 2+ 2a y (15.4-2) dt 2 1 j=1 where: a y, a are the fission product activities of a species S6 2 in volumes 1 and 2, respectively (Curies) Q 12, Q are the transfer rates between the two volumes 21 (ft3 /hr) V7, V2 are the volumes of the unsprayed and sprayed regions of the containment, respectively (ft 3) A A . are the removal coefficients due to the jth 1,j 2,3 removal process in volume 1 and 2, respectively (hrs-1) k y, k 2 are the number of removal processes applicable in volumes 1 and 2, respectively The transfer rate between the sprayed and unsprayed regions was assumed to be limited to the forced convection induced by the fan-cooler units. The flow rate per fan-cooler unit, and the number of units assumed in operation are summarized in table 15.4-11. This assumed minimum flow rate conservatively neglects the effect of natural convection, steam condensation, and diffusion although these effects are expected to enhance the mixing rate between the sprayed and unsprayed volumes. 028 107 0Uh S6 15.4-3 R; s'lu ab POST-CONSTRUCTION PERMIT SUPPLDiENTARY INFORMATION - AUGUST 21, 1979
VNP 15.4.1.3.3 Modeling of Removal Process For fission products other than iodine, the removal processes considered are radioactive decay and leakage from the con-tainment. The decay constants used in the calculations are listed in appendix 15B. The fission product iodine is assumed to be present in the containment atmosphere in elemental, organic, and particulate form. It is assumed that 91 percent of the iodine available for leakage from the containment is in elemental (i.e., I 2 vapor) form, 4 percent is assumed to be in the form of organic iodine compounds (e.g., methyl iodide), and 5 percent is assumed to be adsorbed on airborne particulate matter. For this analysis it was conservatively assumed to subject par-ticulates to a removal rate of 0.5 hr-1 until the particulate iodine in the containment is reduced by a factor of 100 and to subject organic iodines to no removal process other than radioactive decay and leakage from the containment. The effectiveness of the containment spray for the removal of S6 the elemental form of iodine, and the model used to determine the iodine removal efficiency for this application are des-cribed in subsection 6.2.3.3.2 and appendix 6A. The spray iodine removal process is conaidered only in the upper con-tainment volume, which corresponds to the sprayed regions of the containment. A spray removal rate of 20.5 hr-1 is assumed until the airborne elemental iodine in the containment is reduced by a factor of 100. After this time, the spray removal rate is assumed to be zero. The modeling of the unsprayed region (lower containment volume) ds a separate volume eliminates the need to adjust the spray rate constant (As) for the effect of the unsprayed regions. 15.4.1.3.4 Containment Leak Rate The primary containment leak rate used in this analysis is the design-basis leak rate specified in the Technical Specifica-tions. For the first 24 hours following the accident, the leak rate was assumed to be 0.3 percent per day and the leak rate was assumed to be 0.15 percent per day for the remainder of the 30-day period. e MD
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VNP 15.4.1.3.5 Penetration Rooms and Associated Safety Grade Filtration and Exhaust Systems The majority (about 85 percent) of penetrations through the containment wall occur within the electrical and piping pen-etration rooms. These rooms are contained within the fuel handling building, auxiliary building, and control building and are serviced by the safety-grade electrical and piping penetra-tion room filtration and exhaust systems, and the rooms are at a negative pressure when the filtration systems are in operation. All ECCS piping which potentially recirculates contaminated fluids following an accident are completely routed within the containment or the piping penetration room. Any airborne radioactivity released as a result of leakage from these piping systems outside the containment will be collected and filtered prior to release to the environment. Changing the design from 6 the enclosure building to the equipment building and deletion of the enclosure building filtration and exhaust system does not alter the capability to filter potential airborne radio- l activity due to leakage or transfer of post - accident contain-ment fluids. In the calculation of offsite doses, no credit has been taken for filtration of airborne containment leakage even though approximately 85 percent of the containment penetrations are within the negative pressure and filtration boundary. Thus, the majority of the most likely containment liner leak locations are situated in such a manner that filtration of potential airborne radioactivity occurs before release. 15.4.1.3.6 Deleted 15.4.1.3.7 Deleted D IO G3 0 bd CN lh qp r 1[ O .50. k.$ a 9 lP L-9^J 1.v/ S6 15.4-4a POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP Table 15.4-11 PARAMETERS USED IN LOCA ANALYSES REGULATORY GUIDE 1.4 ANALYSIS Core thermal power 3565 MWt 3 Containment free volume 2.75 x 10 6 ft Sprayed containment free volume 2.145 x 10 6 ft 3 3 Unsprayed containment free volume 6.05 x 10 5 fp Primary containment deck fan flow rate 85,000 cfm Number of deck fans assumed operating 4 of 8 Activity released to containment and available for release Noble gases 100% of core inventory
- Iodines 25% of core inventory S6 Form of iodine activity in primary containment available for release Element iodine 91%
Methyl iodine 0I G' D 4% Particulate iodine "U 5%
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GD 1D D 3f Primary containment leak y p -'}b] d 0.30% per day
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3 (0-24 hours) 0.15% per day (1-30 days) Containment spray removal Elemental iodine ~1 20.5 hr Particulates -1 0.5 hr Meteorology See Table 15.B-2
*The activity available for release reflects :he assumption of 50 percent plateout on containment surfaces.
028 110 S6 15.4-5 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP Table 15.4-llA l IODINE AND NOBLE GAS INVENTORY IN REACTOR CORE CORE ACTIVITY ISOTOPE (Curies) 7 I-131 8.80 x 10 I-132 1.34 x 10 8 . 8 1-133 1.97 x 10 8 I-134 2.31 x 10 I-135 1.79 x 10 8 Xe-131m 6.68 x 10 5 56 ye-133 2.03 x 10 8 , Xe-133m 5.16 x 10 6 Xe-135 5.55 x 10 7 Xe-135m 7 5.46 x 10 Xe-138 1.79 x 10 8 Kr-83m 1.64 x 10 , Kr-85 5 9.99 x 10 Kr-85m 3.95 x 10 7 . Kr-87 7.59 x 10 7 Kr-88 1.08 x 10 8 Kr-89 1.40 x 10 8 _ m mA
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NOBLE GAS AND IODINE INVENTORY IN TIIE CONTAINMENT ATMCSPilERE IMFFDIATELY AFTER LOCA AND AVAILABLE FOR LEAKAGE ACTIVITY (Ci) REGULATORY GUIDE ISOTOPE _ _ _ _ 1.4 ANALYSIS 7 I-131 2.20 x 10 I-132'i 2.35 x 10 I-133 4.93 > 10 I-134 5.78 x 10 I 1.'35 4.46 x 10 g S6 Xe-1 1m 6.68 x ?>0 5 8 Xe-133 2.03 x 10 Xe-133m 5.16 x 10 6 7 Xe-135 5.55 x 10 7
'l Xe-135m 5.45 x 13 8
Xe-138 1.79 x 10 7 I ' Kr-83m - 1.64 x 10 . . - Kr-85 ' 9.99 x 10 5 fh((, , j Kr-85m. 3.95 x 10 Kr-87 7.59 x 10 7 Kr-88 1.08 x 10 8 8 Kr-89 1.10 x 10 b b l ! ,2
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S6 15.4-7 POST-CONSTRUCTT.ON PSMIT SUPPLEMENTAPY INFORMATION - AUGUST 21, 1979
VNP 15.4.1.3.8 Results The iodine and noble gas activity releases to atmosphere for the Regulatory Guide 1.4 analysis are given in table 15.4-llc. The gamma and thyroid doses for the loss of coolant accident S6 at the site boundary and the low population zone are given in table it.4-12. The doses are based on the atmospheric dilution
. factors and dose models given in Appendix 15B. The dose limits for this a mident are defined in 10 CFR 100 (25 rem whole body and 300 rem tuyroid). Doses for the Regulatory Guide 1.4 analysis are well within the 10 CFR 100 guidelines.
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e 828 1l3 S6 15.4-8 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
VNP Table 15.4-llc ACTIVITY RELEASES TO ATMOSPHERE FROM LOSS OF COOLANT ACCIDENT Regulatory Guide 1.4 Analysis Activity Release (Ci) ISOTOPE 0-2 HR 2-8 HR 8-24 HR l-4 DAYS 4-30 DAYS I-131 7.59(+2) 9.59(+2) 2.09(+3) 3.99(+3) 1.19(+4) I-132 9.76(+2) 4.01(+2) 6.6(+1) 2.80(-1) 1. 4 (-10) I-133 1.67(+3) 1.87(+3) 2. 96 (+3) 1.91(+3) 1.93(+2) I-134 1.35(+3) 1.22(+2) 8.30(-1) 1.2(-6) 2. l (-31) I-135 1.45(+3) 1.22(+3) 9.72(+2) 1.14 (+2 ) 6.8(-2) S6 Xe-133 5.06(+4) 1.48(+5) 3. 72 (+5) 6. 60 (+5) 1. 31 (+ 6 ) Xe-133m 1.28(+3) 3.64(+3) 8. 4 3 (+3) 1.11(+4) 7.30(+3) Xe-135 1.28(+4) 2.86(+4) 3. 50 (+4 ) 7.40(+3) 3.22(+1) Xe-135m 2.56(+3) 1.24(+1) 1.39(-6) 2.00(-28) NEGLIGIBLE Xe-138 9.07(+3) 6.80(+1) 2.80(-5) 1.30(-22) NEGLIBIBLE tr-85 4.22(+2) 1.26(+3) 3. 37 (+3) 7.55(+3) 6.39(+4) Kr-8mn 1.d3(+4) 2.34(+4) 1.35(+4) 5.75(+2) 6.10(-3) Kr-87 1.94(+4) 9.43(+3) 3. 68 (+2) 2.90(-2) 2.27(-19) Kr-86 3.60(+4) 4.30(+4) 1. 21 (+4 ) 1.13(+2) 1.71(-6) Note: 6. 21 (-5) = 6.21 x 10-5 g, vu uu
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o n Table 15.4-12 to POTENTIAL OFFSITE DOSES DUE TO ACCIDENTS E VOGTLE NUCLEAR PLANT 3 o Z DOSE (2 HOURS) AT DOSE (COURSE OF y EXCLUSION AREA ACCIDENT) AT LOW g BOUNDARY POPULATION ZONE s e (1060 Meters) (3218) meters) S6 PSAR Thyroid Whole Body Thyroid Whole Body y Postulated Section (rem) (rem) (rem) (rem) M gg LOCA 15.4 >m N @* Containment Leak *
$ Conservative 93.3 7.43(+0) 103 5.47(+0) o a
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VNP TABLE OF CONTENTS APPENDIX 15B Section Title Page 15B DOSE MODELS USED TO EVALUATE THE ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15B.1 INTRODUCTION S6 15B-1 15B.2 ASSUMPTIONS D
' rN S6 15B-1 D
15B.3 GAMMA DOSE Cs L- S6 15B-1 S6 15B.4 THYROID INHALATION DOSE t A E 6 15B-2 C) -< .A a 15B.5 REFERENCES S6 15B-3 l 15B DOSE MODELS USED TO EVALUATE THE ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15B.1 INTRODUCTION 15B-1 15B.2 ASSUMPTIONS 15B-1 15B.3 GAMMA AND BETA DOSE 15E-2 15B.4 THYROID INHALATION DOSE 15B-3 15B.5 REFERENCES 15B-6 LIST OF TABLES Table Title Page , 15B-1 PHYSICAL DATA FOR ISOTOPES S6 15B-4 15B-2 ACCIDENT ATMOSPHERIC DILUTION FACTORS (X/Q) AT EXCLUSION AREA BOUNDARY AND LOW POPULATION ZONE FOR THE VOGTLE NUCLEAR PLANT 56 15B-5 15B-1 PHYSICAL DATA FOR ISOTOPES 15B-4 15B-2 ACCIDENT ATMOSPHERIC DILUTION FACTORS 15B-5 b b llb S6 15B-i ._ > v - , u , .. , ,
VNP APPENDIX 15B DOSE MODELS USED TO ZVALUATE THE ENVIRONMENTAL CONSEQUENCE 2 OF ACCIDENTS 15B.1 INTRODUCTION This section identifies the models used to calculate the off-site radiological doses that would result from releases of radioactivity due to various postulated accidents. lS6 15B.2 ASSUMPTIONS The following assumptions are basic to both the model for the gamma doses due to immersion in a semi-infinite cloud of radio- lS6 activity, and the model for the thyroid dose due to inhalation of radiciodine: lS6
- a. Direct radiation from the source point is negligible compared to gamma radiation due to immersion in the S6 semi-infinite radioactive cloud.
- b. All radioactivity releases are treated as ground level releases regardless of the point of discharge.
- c. The dose receptor is a standard man, as defined by lS6 the International Commission on Radiological Protec-tion (ICRP) (Reference 1).
- d. Radioactive decay from the point of release to the dose receptor is neglected.
- e. Isotopic data such as decay ra s and decay energy emissions are taken from Refertace 2.
15B.3 GAMMA DOSE T gamma dose delivered to a dose receptor is obtained by con-su cing the dose receptor to be immersed in a radioactive cloud which is infinite in all directions above the ground plane, i.e., an " infinite semi-spherical cloud." The concentration of radioactive material within this cloud is taken to be uniform, and equal to the maximum centerline ground level concentration that would exist in the cloud at the appropriate distance from the point of release. D lh u eu JL
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S6 15B-1 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, '979
VNP The gamma dose is a result of external gamma radiation. Equations describing a semi-infinite cloud were used to calcu-late the doses for a given time period as follows: (Reference 3). Gamma Dose = 0.25 h- R f g i (15B-1) where: S6 A is the activity of isotope i released during a given R i time period (Curies) X_ is the atmospheric dilution factor for a given time Q period (table 15B-2) (sec/m3) is the average gamma radiation energy emitted by E isotope i per disintegration (meV/ disintegration)
}i 15B.4 THYROID INHALATION DOSE The thyroid dose for a given time period t is obtained from the following expression: (Reference 4)
IQ D = X/O B - 11 DCF f (15B-2) where: D = thyroid inhalation dose, rem (X/Q) t
= site dispersion factor for time interval t, sec/m3 B = brec ..g rate for time interval t, m3/sec Q.1 = total activity of iodine isotope i released in time period t, Curies (DCF)1
. = dose conversion factor for iodine isotope i, rem / Curie inhaled The isotopic data and " standard man" data are given in table S6 15B-1. The gamma energies, E y, on table 15B-1 include the X-rays and annihilation gamma rays if they are prominent in the electromagnetic spectrum.
,e f[ e 1 hJJL 3 S6 5B-2 828 18 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 197S
VNP 15B.5 REFERENCES
- 1. " Report of ICRP Committee II on Permissive Dose for Internal Radiation (1959) ," Health Physics Volume 3,, pp. 30, 146-153, 1970. lS6
- 2. Lederer, C. M., Hollander, J. M., Perlman, I. " Table of Isotopes, Sixth Edition".
- 3. Regulatory Guide 1.4, Revision 2, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors,"
USNRC, June 1974. S6
- 4. Regulatory Guide 1.109, Revision 1, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," USNRC, October 1977.
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C 3 119 S6 15B-3 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
Vh7 6 Table 15B-1 PHYSICAL DATA FOR ISOTOPES DOSE DECAY GAMMA CONVERSION CONSTANT ** ENERGY ** FACTOR
- ISOTOPE (hr-1) (Mev/Disint.) (rem / Curie) 6 I-131 3.587 x 10 -3 0.371 1.49 x 10
-1 4 I-132 3.01 x 10 2.400 1.43 x 10
-2 5 I-133 3.33 x 10 0.477 2.69 x 10 I-134 7.87 x 10 -1 1.94 3.73 x 10 3
-1 4 I-135 1.04 x 10 1.78 5.6 x 10
-3 Xe-133 5.48 x 10 0.03 -
-2 Xe-133m 1.28 x 10 0.0326 -
Xe-135 7.53 x 10" 0.246 - Xe-135m 2.67 x 10 0 0.422 - 0 Xe-138 2.45 x 10 2.87 -
-6 Kr-85 7.35 x 10 0.00211 -
-1 Kr-85m 1.59 x 10 0.151 -
-1 Kr-87 5.47 x 10 1.37 -
-1 Kr-88 2.50 x 10 1.74 -
BREATHING RATES Time Period Breathing Rates O bD Q (Hours) Ig U] O U (m3/ sec) 0-8 p [ 3.47 x 10 -4 8 24 k 1.'i5 x 10
-4 24 - 720 ^ 32 x 10
- Referer.ce 4
** Reference 2 S6 15B-4 "B
'~ 1'30 POS'.'-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 2'. , 1979
VNP Table 15B-2 ACCIDENT ATMOSPHERI.C DILUTION FACTORS (X/Q)* AT EXCLUSION AREA BOUNDARY AND LOW POPULATION ZONE FOR THE VOGTLE NUCLEAR PLANT EXCLUSION AREA BOUNDARY LOW POPULATION ZONE TIME PERIOD (1060 Meters) (3218 Meters) (hours) S6 0-2 1.6(-4) 6. 3 (-5) 0-8 3.2(-5) 8-24 2.2(-5) 24-96 1. 0 (-5 ) 96-720 3.3(-6) D n? cN 0 cv av C3 I)) D~l ~ o fu ' J)}1 1{ ^ 2 X/O values, expressed in sec/m3 , are based on the guidelines set forth in the August 1978 draft of Regulatory Guide 1.XXX. The site data used was taken during the recent year (1977-1978). S6 NOTE: 1.6(-4) = 1.6 x 10 -4 8 3 12i S6 15B-5 p - ) v T _ o
VNP Table 15B-2 ACCIDENT ATMOSPHERIC DILUTION FACTORS (X/Q)
- AT EXCLUSION AREA BOUNDARY AND LOW POPULATION ZONE FOR THE VOGTLE NUCLEAR PLANT EXCLUSION AREA BOUNDARY LOW POPULATION ZONE TIME PERIOD (1060 Meters) (3218 Meters)
(hours) S6 0-2 1.6(-4) 6.3(-5) 0-8 3.2(-5) 8-24 2.2(-5) 24-96 1.0(-5) 96-720 3.3(-6) D 07 CN D cV cu JU
- X/O values, expressed in sec/m3 , are based on the guidelines set forth in the August 1978 draft of Regulatory Guide 1.XXX.
The site data used was taken during the recent year (1977-1978). S6 NOTE: 1.6(-4) = 1.6 x 10 -4 878 122 S6 15B-5 POST-CONSTRUCTION PERMIT SUPPLEMENTAR - 1 1
VNP TABLE OF CONTENTS (Continued) Section Title Page 3.1 REACTOR COOLANT SYSTEM 16.3.1-1 3.1.1 APPLICABILITY 16.3.1-1 3.1.2 OBJECTIVE 16.3.1-1 3.1.3 SPECIFICATION 16.3.1-1 3.1.3.1 Operational Components 16.3.1-1 3.1.3.2 Heatup and Cooldown 16.3.1-3 3.1.3.3 Minimum conditions for Criticality 16.3.1-8 3.1.3.4 Maximum Reactor Coolant Activity 16.3.1-9 3.1.3.5 Leakage 16.3.1-11 3.1.3.6 Maximum Reactor Coolant Oxygen, Chloride, and Fluoride Concentration 16.3.1-13 3.2 CHEMICAL AND VOLUME CONTROL SYSTEM 16.3.2-1 3.2.1 APPLICABILITY 16.3.2-1 3.2.2 OBJECTIVE 16.3.2-1 3.2.3 SPECIFICATION 16.3.2-1 3.2.4 BASIS 16.3.2-2 3.2.5 REFERENCE 16.3.2-3 3.3 ENGINEERED SAFETY FEATURES 16.3.3-1 3.3.1 APPLICABILITY 16.3.3-1 3.3.2 OBJECTIVE 16.3.3-1 3.3.3 SPECIFICATION 16.3.3-1 3.3.3.1 Emergency Core Cooling Systems 16.3.3-1 023 123 B 'O h I S6 16-iii GU T 21 1979
VNP TABLE OF CONTENTS (Continued) Section Title Page 3.3.3.2 Containment Cooling and Iodine Removal Systems 16.3.3-3 S6l3.3.3.3 Deleted S6 16.3.3-4 3.3.3.3 Enclosure Building Filtration System 16.3.3-4 3.3.3.4 Penetration Room Filtration System 16.3.3-5 3.3.3.5 Component Cooling System 16.3.3-5 3.3.3.6 Nuclear Service Water System 16.3.3-6 3.3.3.7 Ultimate Heat Sink 16.3.3-7 3.3.3.8 Post-Accident Containment Venting System 16.3.3-8 3.3.3.9 Post-Accident Hydrogen Control System 16.3.3-8 3.3.4 BASIS 16.3.3-9 3.
3.5 REFERENCES
16.3.3-12 3.4 SECONDARY STEAM AND POWER CONVERSION SYSTEM 16.3.4-1 3.4.1 APPLICABILITY 16.3.4-1 3.4.2 OBJECTIVE 16.3.4-1 3.4.3 SPECIFICATION 16.3.4-1 3.4.4 BASIS 16.3.4-1 3.5 INSTRUMENTATION SYSTEM - OPERATIONAL SAFETY INSTRUMENTATIO_N_ 16.3.5-1 3.5.1 APPLICABILITY 16.3.5-1 1.5.2 OBJECTIVES O' D 16.3.5-1 3.5.3 SPECIFICATION dbo 16.3.5-1 3.5.4 BASIS 0 $ 16.3.5-7 3.
5.5 REFERENCES
16.3.5-11 3.6 CONTAINMENT SYSTEMS 16.3.6-1 023 124 S6 16-iv POST-CON TRUCTION PERMIT .UPP TAR ' FORMAT O - AUGUST 21 1979
VNP TABLE OF CONTENTS (Contin';al Section Title Paga 4.12.4 BASIS 16.4.12-2 5 DESIGN FEATURES 16.5.1-1 5.1 SITE 16.5.1-1 5.2 CONTAINMENT 16.5.2-1 5.2.1 CONTAINMENT 16.5.2-1 5.2.1.1 Shielding S6 16.5.2-1 lS6 5.2.1.2 Internal Pressure, Temperature, and Vacuum 16.5.2-1 5.2.2 DELETED S6 16.5.2-1 lS6 5.2.2 ENCLOSURE BUILDING 16.5.2-1 5.2.3 PENETRATION ROOM 16.5.2-1 5.2.4 PENETRATIONS 16.5.2-1 5.2.4.1 Seals 16.5.2-2 5.2.4.2 Leakage Barriers 16.5.2-2 5.2.5 POST LOCA CONTAINMENT PRESSURE SUPPRESSION SYSTEMS 16.5.2-2 5.2.5.1 Internal Spray Systems 16.5.2-2 5.2.5.2 Cooling Systems 16.5.2-2 5.2.5.3 Air Filter Units c3 c3 16.5.2-3 5.3 REACTOR e, e, 16.5.3-1 5.3.1 APPLICABILITY 16.5.3-1 A 5.3.2 OBJECTIVE 16.5.3-1 5.3.3 SPECIFICATION 16.5.3-1 5.3.3.1 Reactor Core 16.5.3-1 5.3.3.2 Reactor Coolant System 16.5.3-2 rn ,,-
' ','1 S6 16-xi
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VNP TABLE OF CONTENTS (Continued) Sy tion Title Page 5.4 FUEL STORAGE 16.5.4-1 5.4.1 FUEL STRUCTURES 16.5.4-1 5.4.2 FUEL STORAGE RACKS 16.5.4-1 5.4.3 SPENT FUEL STORAGE POOL 16.5.4-1 6 ADMINISTRATIVE CONTROLS 16.6.1-1 6.1 ORGANIZATION, REVIEW AND AUDIT 16.6.1-1 6.
1.1 INTRODUCTION
16.6.1-1 6.1.2 PLAN'. SUPERINTENDENT RESPONSIBILITY 16.6.1-1 6.1.3 ORGP'.IZATION FOR CONDUCT OF OPERATION 16.6.1-1 6.1.4 ORGANIZATIONAL UNITS 16.6.1-1 6.1.4.1 Plant Review Board (PRB) 16.6.1-3 6.1.4.2 Safety Review Board (SRB) 16.6.1-5 6.2 ACTION TO BE TAKEN IN THE EVENT OF AN ABNORMAL OCCURRENCE IN PLANT OPERATION 16.6.2-1 6.3 ACTION TO BE TAKEN IF A SAFETY LIMIT IS EXCEEDED 16.6.3-1 6.4 UNIT OPERATING PROCEDURES 16.6.4-1 6.4.1 SYSTEMS AND COMPONENTS INVOLVING NUCLEAR SAFETY 16.6.4-1 6.4.2 RADIATION CONTROL 16.6.4-1 6.5 UNIT OPERATING RECORDS 16.6.5-1 6.6 PLANT REPORTING REQUIREMENTS 16.6.6-1 6.6.1 ROUTINE REPORTS 16.6.6-1 6.6.2 NONROUTINE REPORTS D 16.6.6-8 0930 mm S6 16-xii POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21 1979
TECHNICAL SPECIFICATIONS VNP
- 5. One residual heat exchanger may be out of service provided that it is restored to operable status within (FSAR) days. If one pump is out of service per the previous paragraph, it must be the one in the same train as the out-of-service heat exchanger.
- 6. Any valve required for the functioning of the system during and following accident conditions may be inoperable provided it is restored to operable status within (FSAR) hours and all valves in the system that provide duplicate function are demonstrated to be operable.
- 7. During normal operation, one boron injection tank heater channel and/or one channel of line heat tracing may be inoperable for an extended period.
3.3.3.2 Containment Cooling and Iodine Removal Systems A. The reactor shall not be made critical, except for low power physics tests, unless the following conditions are met:
- 1. The spray additive tank contains not less than 3200 gallons of solution with hydroxide concentration of not less than 30 percent by weight.
- 2. Two containment spray trains, including containment spray pumps, piping, and valves shall be operable.
- 3. Four fan cooler units including all essential valves, controls, dampers, and piping associated with these units shall be operable.
B. During power operation the requirements of speci-fication 3.3.3.2.1 may be modified to allow the following components to be inoperable. If the components are not restored to meet the requirements of specification 3.3.3.2.1 within the time period specified below, the reactor shall be placed in the hot shutdown condition. If the requirements of specification 3.3.3.2.1 are not satisfied within an additional 48 hours, the reactor shall be placed in D f D
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TECHNICAL SPECIFICATIONS VNP the cold shutdcwn condition using normal operating procedures.
- 1. One containment spray train may be out of service, provided immediate attention is directed to making repairs and the train can be restored to operable status within (FSAR) hours. The other containment spray train shall be tested as specified in specification 4.5 to demonstrate operability prior to initiating repair of the inoperable system, or
- 2. One containment fan cooler unit may be out of service, provided immediate attention is directed to making repairs and the fan cooler unit can be restored to operable status within (FSAR) hours.
Two of the remaining fan cooler units shall be tested as specified in specification 4.5 to demonstrate operability prior to initiating repair of the inoperable unit. S6l 3.3.3.3 Deleted e ogg a3d@ e e 823 128 S6 16.3.3-4 POS'c-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
TECHNICAL SPECIFICATIONS VNP
- 2. Acceptable Criteria e a. System tests will be considered satisfactory if visual observations indicate all compo-nants of each system have functioned properly,
- b. Containment spray pump tests will be co.,3idered satisfactory at miniflow condition
@. if the pumps reach their maximum required flow. The discharge pressure and corresponding flow rate determine a point on the head curve; and the pumps operate for at Jeast fifteen minutes.
- c. A test of a motor operated valve shall be considered satisfactory if its limit switch operates a light on the main control board demonstrating that the valve has stroked.
- d. The test of the containment spray nozzles shall be considered satisfactory if air flow or smoke through the nozzles indicates that the nondes are not plugged.
This space has been left o~ lank intentionally lS6 1 g g 10 1RJ o
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P98 129 S6 16.4.5-3 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21 1979
TECHNICAL SPECIFICATIONS VNP E. Penetration Room Filtration System System tests shall be made at (FSAR) intervals while the unit is operating. These tests shall consist of visual inspection, a flow measurement using the pitot tube installed at the outlet of each filter subsystem, and pressure drop measurements across each filter bank. Visual inspection shall include inspection of general conditions for evidence of: water, oil, or other foreign material; gasket deterioration; and adhesive deterioration in the HEPA units. These tests will be performed within (FSAR) hours after startup from cold shutdown. The test will be considered satisfactory if visual inspection reveals no abnormal conditions, flow is equal to design flow or higher, no unusual or excessive noise or vibration exists when the fan motors are operating, and the pressure drop across any filter bank does not exceed two times the pressure drop when new. 4.5.3.2 Component Tests A. Pumps
- 1. Safety injection pumps, residual heat removal pumps, containment spray pumps, and the auxiliary component cooling water pumps shall be started at intervals not greater than (FSAR).
- 2. Acceptable levels of performance shall be that the pumps start, reach their required developed head on recirculation flow, and operate for at least (FSAR) minutes.
B. Valves
- 1. Each boron injection tank outlet valve shall be cycled by operator action with the pumps shut down at intervals not greater than once every refueling.
- 2. Each spray additive valve shall be cycled by operator action with tl:e pumps shut down at inv arvals not greater than once every refueling.
898 130 O f? GD c) c) 9 O lb f}$ If S6 16.4.5-4 d - -
.13 POS'/-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
TECHNICAL SPECIFICATIONS VNP 5.2 CONTAINMENT 5.2.1 CONTAINMENT 5.2.1.1 Shielding lS6 The containment completely encloses the reactor and reactor coolant syster.and ensures that an acceptable upper limit for leakage of radioactive materials to the environment is not exceeded even if gross failure of the reactor coolant system occurs. The structure provides biological shielding for both normal and accident situations. 5.2.1.2 Internal Pressure, Temperature, and Vacuum The containment is designed for an internal pressure of 47 psig and an internal temperature of 271 F. The containment is also structurally designed to withstand an internal vacuum of 3.0 psig. 5.2.2 DELETED S6 This space has been left blank intentionally. 5.2.3 PENETRATION ROOM The penetration room filtration system collects, controls, and minimizes the release of radioactive materials from the containment to the environment following a design basis accident. Air from the penetration rooms is drawn through the filter subsystem consisting of prefilters, high efficiency particulate filters, and charcoal filters in service. The air is then discharged through the vent. The system provides for enhanced dispersion of fission products by mixing of contain-ment penetration leakage with all areas of the penetration D N r> D cV cu id
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9d l{ 6 - - 3 0^: 171 S6 16.5.2-1 U :'_ < iJl POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979
TECHNICAL SPECIFICATIONS VNP rooms, increased holdup time within the penetration rooms, improved cleanup of fission products by filtration, and a slightly negative pressure in the penetration rooms to insure inleakage. 5.2.4 PENETRATIONS 5.2.4.1 Seals Penetration assemblies are seal welded to the containment liner. Access openings, electrical penetration canisters, and fuel transfer tube covers are equipped with double seals. Containment purge penetrations and containment atmosphere sampling penetrations are equipped with double valves having resilient seating surface. 5.2.4.2 Leakage Barriers Leakage through all fluid penetrations not serving accident-consequence-limiting systems is minimized by a double barrier so that no single credible failure or malfunction of an active component can result in loss-of-isolation or intolerable leakage. The installed double barriers take the form of closed piping systems, both inside and outside.the containment, and various types of isolation valves. 5.2.5 POST LOCA CONTAINMENT PRESSURE SUPPRESSION SYSTEMS 5.2.5.1 Internal Spray Systems The containment has redundant internal spray systems capable of providing a distributed borated water spray of at least 2,600 gal./ min. During the initial period of spray operation, sodium hydroxide is added to the spray water to increase the removal of iodine from the containment atmosphere. 5.2.5.2 Cooling Systems The containment has redundant containment cooling systems, which consist of four containment air coolers (each consisting of a fan and a water-cooled heat exchanger) with total heat removal capability of twice the design requirement under conditions following LOCA. f ye h O 7 q ') O) L' k )9w S6 16.5.2-2 POST-CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - AUGUST 21, 1979}}