ML022210409
| ML022210409 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 07/30/2002 |
| From: | Mecredy R Rochester Gas & Electric Corp |
| To: | Jack Cushing Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML022210409 (197) | |
Text
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Volume 1 Table of Contents Section 1 - Administrative Information Section 2 - Scoping & Screening Methodology/Results
APPLICATION FOR RENEWED OPERATING LICENSE R. E. GINNA NUCLEAR POWER PLANT
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information TECHNICAL AND ADMINISTRATIVE INFORMATION CONTENTS Administrative Information ---- ---------------------------------------- 1-1 1.0 1.1 Application Format and Content ---- ------------------------------------- 1-2 Plant Description ----- ----------------------------------------------- 1-4 1.2 Information Required by 10 CFR 54.17 and 10 CFR 54.19 -- --------------------- 1-4 1.3 1.3.1 Name of Applicant ---- ---------------------------------------- 1-4 1.3.2 Address of Applicant ---- --------------------------------------- 1-4 1.3.3 Description of Business or Occupation of Applicant -- ------------------- 1-5 1.3.4 Organization and Management of Applicant --- ----------------------- 1-5 1.3.5 Class of License, Use of Facility, and Period of Time for which the License is Sought 1-9 1.3.6 Earliest and Latest Dates for Alterations, if Proposed -- ------------------ 1-9 1.3.7 Listing of Regulatory Agencies Having Jurisdiction and News Publications - ------ 1-10 1.3.8 Conforming Changes to Standard Indemnity Agreement -- ---------------- 1-10 1.3.9 Restricted Data Agreement ---- ---------------------------------- 1-10 Current Licensing Basis Changes During NRC Review --- ---------------------- 1-11 1.4 Abbreviations and Passive Function Code Definitions -- ------------------------ 1-12 1.5 1.5.1 Abbreviations .........-----.. -.---- --------------------------
..- 1-12 1.5.2 Passive Function Code Definitions --- ------------------------------ 1-19 Communications ----- ----------------------------------------------- 1-24 1.6 1-25 Section 1.0 References ----- -----------------------------------------------
2.0 Scoping and Screening Methodology for Identifying Structures and Components Subject to Aging Management Review, and Implementation Results -- -------------------- 2-1 Scoping and Screening Methodology ---- ---------------------------------- 2-1 2.1 Introduction .........-----.. ..-
-.- -.---------------------------- 2-1 2.1.1 2.1.2 Plant Level Scoping ---- --------------------------------------- 2-1 System and Structure Function Determination -- ----------------------- 2-2 2.1.3 2.1.4 Design Codes, Standards, and SSC Safety Classifications - --------------- 2-5 2.1.5 Application of License Renewal Scoping Criterion -- -------------------- 2-6 2.1.5.1 Safety-Related Criteria Pursuant to 10 CFR 54.4(a)(1) (Criterion 1) - - - 2-6 2.1.5.2 Non-Safety Related Criteria Pursuant to 10 CFR 54.4(a)(2) (Criterion 2) - - 2-7 2.1.5.3 Phase 1: Incorporating the Non-Safety SSCs Meeting the Criterion that are Explicitly Identified in the Current Licensing Basis - -------------- 2-7 Page i
Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information TECHNICAL AND ADMINISTRATIVE INFORMATION Contents 2.1.5.4 Phase 2: Evaluation of Non-Safety SSCs for Failure Modes and Effects or Spatial Interactions not Explicitly Functionally Described in the CLB - -------- 2-11 2.1.5.5 Other Scoping Pursuant to 10 CFR 54.4(a)(3) (Criterion 3) - - - - - - 12 2.1.5.6 Fire Protection (FP) .....-.----.-.-.------------- 2-12
- - - - 13 2.1.5.7 Environmental Qualification (EQ)--------------.-.--.
Pressurized Thermal Shock (PTS)--------------.--.- ----- 2-13 2.1.5.8 2.1.5.9 Anticipated Transients Without a Scram (ATWS) - - .- - - - - - 2-13 2.1.5.10 Station Blackout (SBO)--------------------.-.-.-- ----- 2-14 2.1.6 Interim Staff Guidance Discussion .....----.-.-.------------- ----- 2-14 2.1.7 Component Level Screening (Identification of Components Subject to Aging Management R e v ie w ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - - - - 18 2.1.7.1 Mechanical Systems ---------.-.--. - - - - - - - - - - - - - 19
- - - - - - - - - 2-20 2.1.7.2 Civil Structures --------------.--.
Structural Commodity Groups ----.-.- - - - - - - - - - - - - -- 2-22 2.1.7.3 Electrical and I&C Systems --.-----.- - - - - - - - - - - - - - 22 2.1.7.4 Screening of Stored Equipment -----.- - - - - - - - - - . . . 24 2.1.7.5 2.1.7.6 Screening of Thermal Insulation ---.-.- - - - - - - - - - - - - - - 25 2.1.7.7 Identification of Short-lived Components and %.JVII*-1
2-25
. . . . . . . . . . . . . . . . . . . . - - - -- 2-27 Section 2.1 References
- - - -- 2-34 2.2 Plant Level Scoping Results ----------------.-.-.--
- - - - 40 2.3 Scoping and Screening Results: Mechanical Systems ---.-
- -- - - 2-40 2.3.1 Reactor Coolant System--------------.--.-
- -. 2-40 2.3.1.1 Reactor Coolant (Class 1)-------.--.
2.3.1.2 Reactor Vessel--------------.--. 2-46 Reactor Vessel Internals--------.--. - - - -- 2-50 2.3.1.3 Pressurizer ---------------.-.--. 2-54 2.3.1.4 Steam Generators ------------.--. 2-56 2.3.1.5
- - - -- 2-59 2.3.1.6 Reactor Coolant (Non-Class 1) -----.-
- - - -- 2-65 2.3.2 Engineered Safety Features Systems --.-----.-
- - - -- 2-65 2.3.2.1 Safety Injection (SI) ---------.-.--.
- - - 2-72 2.3.2.2 Containment Spray (CS)--------.--.
- - - -- 2-77 2.3.2.3 Residual Heat Removal (RHR) -
2.3.2.4 Containment Hydrogen Detectors and Recombineers-----. -- - - - -- 2-83 2.3.2.5 Containment Isolation Components ....... - - - -- 2-88 Page-ii
Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information TECHNICAL AND ADMINISTRATIVE INFORMATION Contents 2.3.3 Auxiliary SSystems ......- -.- -.-.------------------- - - 2-92 2.3.3.1 Chemical and Volume Control (CVCS)-------.--. - - 2-92 2.3.3.2 Component Cooling Water (CCW)------------.- - - 2-100 2.3.3.3 Spent Fuel Cooling and Fuel Storage ---.-----.- - - 2-104 2.3.3.4 Waste Disposal .....----.-.-.------------- - - 2-112 2.3.3.5 Service Water (SW) ....----.-.------------. - 117 2.3.3.6 Fire Protection (FP) ....----.-.------------. - - 2-122 2.3.3.7 Heating Steam --------------------.-.-.-- - - 2-134 2.3.3.8 Emergency Power-------------------.-.--. - - 2-137 2.3.3.9 Containment Ventilation ---------------.-.--. - - 2-143 2.3.3.10 Essential Ventilation Systems ----------.-.-.-- - - 2-150 2.3.3.11 Cranes, Hoists, and Lifting Devices -----------.- - - 2-159 2.3.3.12 Treated Water .....-----.-.-.------------- -- 2-161 2.3.3.13 Radiation Monitoring ----------------.-.-.-- - - 2-165 2.3.3.14 Circulating Water - Not Within Scope of License Renewal 2-170 2.3.3.15 Chilled Water - Not Within Scope of License Renewal - - - 2-171 2.3.3.16 Fuel Handling - Not Within Scope of License Renewal- 2-172 2.3.3.17 Plant Sampling - Not Within Scope of License Renewal - - - 2-174 2.3.3.18 Plant Air - Not Within Scope of License Renewal ----.- - - 2-177 2.3.3.19 Non-Essential Ventilation - Not Within Scope of License Ren4ewal - - 2-179 2.3.3.20 Site Service and Facility Support - Not Within Scope of Licen se Renewa 1 181 2.3.4 Steam and Power Conversion Systems -.--------...... - 183 2.3.4.1 Main and Auxiliary Steam ---- ----------------------------- 2-183 2.3.4.2 Feedwater and Condensate --- ---------------------------- 2-189 2.3.4.3 Auxiliary Feedwater (AFW) --- ---------------------------- 2-195 2.3.4.4 Turbine-Generator and Supporting Systems -- ----------------- 2-201 2.4 Scoping and Screening Results: Structures --- ------------------------------- 2-205 2.4.1 Containment Structures ---- ------------------------------------- 2-205 2.4.2 Essential Buildings and Yard Structures --- --------------------------- 2-220 2.4.2.1 Auxiliary Building ---- ----------------------------------- 2-221 2.4.2.2 Intermediate Building ---- -------------------------------- 2-233 2.4.2.3 Turbine Building ---- ----------------------------------- 2-244 2.4.2.4 Diesel Building ---- ------------------------------------ 2-254 2.4.2.5 Control Building----------------.-.--. - 262 Page-iii
Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information TECHNICAL AND ADMINISTRATIVE INFORMATION Contents 2.4.2.6 All Volatile Water Treatment Building --- --------------------- 2-274 2.4.2.7 Screen House Building ---- ------------------------------ 2-282 2.4.2.8 Standby Auxiliary Feedwater Building --- --------------------- 2-291 2.4.2.9 Service Building ---- ----------------------------------- 2-298 2.4.2.10 Cable Tunnel ---- ------------------------------------- 2-306 2.4.2.11 Essential Yard Structures ---- ---------------------------- 2-312 2.4.2.12 Component Supports Commodity Group -- -------------------- 2-315 2.4.3 Non-Essential Buildings and Yard Structures - Not within Scope of License Renewal- 2-324 2.5 Screening Results: Electrical and Instrumentation and Controls Systems - ------------ 2-327 2.5.1 Commodity Group Descriptions ---- -------------------------------- 2-327 2.5.2 120 VAC Vital Instrument Buses - Within Scope of License Renewal- --------- 2-330 2.5.3 125 VDC Power - Within Scope of License Renewal -- ------------------- 2-333 2.5.4 4160 VAC Power- Within Scope of License Renewal -- ------------------- 2-337 2.5.5 480 VAC Power - Within Scope of License Renewal -- ------------------- 2-339 2.5.6 Control Rod Drive and Nuclear Process Instrumentation - Within Scope of License Renewal ----- ----------------------------------------------- 2-344 2.5.7 Misc. AC Power and Lighting - Within Scope of License Renewal - ---------- 2-349 2.5.8 Offsite Power - Within Scope of License Renewal --- -------------------- 2-352 2.5.9 Reactor Protection - Within Scope of License Renewal -- ----------------- 2-355 2.5.10 Engineered Safety Features Actuation System - Within Scope of License Renewal- 360 2.5.11 Plant Communications - Within Scope of License Renewal -- -------------- 2-363 2.5.12 Plant Process Computer - Not Within Scope of License Renewal - ----------- 2-366 2.5.13 Plant Security - Not Within Scope of License Renewal -- ------------------ 2-368 2.5.14 Seismic and Meteorological Instrumentation System- Not Within Scope of License Renewal ----- ----------------------------------------------- 2-370 3.0 Aging Management Review Results --- ----------------------------------- 3-1 3.1 Review Methodology ---- --------------------------------------------- 3-1 3.1.1 Determination of Materials of Construction -- -------------------------- 3-1 3.1.2 Determination of Operating Environments --- ------------------------- 3-1 3.1.3 Component Grouping by MateriaVEnvironment Combination - -------------- 3-2 3.1.4 Aging Effects Analysis - Non-Class 1 Mechanical Systems and Components - ----- 3-2 Page-iv
Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information TECHNICAL AND ADMINISTRATIVE INFORMATION Contents 3.1.4.1 Treated Water Systems .......- -.- -.--------------------- 3-2 3-2 3 .1.4 .2 P rotective C oatings . . . . . . . . . . . . . . . . . . . . . .
3-3 3.1.5 Aging Effect Analysis - Essential Structures--------------------.-.-.--
3-3 3.1.6 Aging Effects Analysis - Class 1 Systems, Structures and Components ----.-.-
3-3 3.1.7 Industry and Plant-Specific Operating Experience Review ----------.-.-.--
3-4 3.1.8 Assignment of Aging Management Programs --------------------.-.--.
3-4 3.1.9 Standard Industry Guidance Document Review (Mechanical Systems and Components) 3-4 3.1.9.1 Position on SCC of Austenitic Stainless Steel --------------.--.-
3-5 3.1.9.2 Position on IGA/IGSCC of Austenitic Stainless Steels -----------.-
3-5 3.1.9.3 Position on SCC of SA 193 Grade B7 Bolting Materials --------.--.
3-6 3.1.10 Standard Industry Guidance Document Review (Structures and Structural Components) 3.1.11 Standard Industry Guidance Document Review (Electrical Commodities) - 3-6 3-6 3.1.12 Generic Component Assets-----------------------.--.-
3-7 3.1.13 Review of NUREG-0933 ......-.--.-.-.----------------
- - -- - - - 3-12 Se c tion 3 .1 R e fe re n c e s . . . . . . . . . . . . . . . . . . . . . . . . . . . .
- 13 3.2 Aging Management of Reactor Coolant System---------------.-.--.
3-76 S e c tio n 3 .2 R e fe re n c e s . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3-77 3.3 Aging Management of Engineered Safety Features Systems -.-.-.------
3-99 3.4 Aging Management of Auxiliary Systems --------------------.-.--.
- 194 3.5 Aging Management of Steam and Power Conversion Systems --.-----.-
- 214 3.6 Aging Management of Structures and Component Supports--------.--.
- 296 Se ctio n 3 .6 R e fe re n ce s . . . . . . . . . . . . . . . . . . . . . . . . . . . .
- 297 3.7 Aging Management of Electrical and Instrument and Controls Systems ----
- 309 S ection 3.7 R eferences . . . . . . . . . . . . . . . . . . . . . . . . . . . .
4-1 4.0 Tim e-Lim ited Aging Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1 Identification of Time-Limited Aging Analyses ......-.--.-.-.----------------- 4-1 4-1 4.1.1 Identification Process of Time-Limited Aging Analyses --------------.-.--.
4-2 4.1.2 TLAA Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
4-5 4.1.3 Identification and Evaluation of Active Plant-specific Exemptions ----.-----.-
4-5 4.1.4 Screening of Potential Time-Limited Aging Analyses ---------------.-.--.-
Page-v
Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information TECHNICAL AND ADMINISTRATIVE INFORMATION Contents 4.1.5 Evaluation Process of Time-Limited Aging Analyses - - -.- - - ---- 4-6 4.2 Reactor Vessel Neutron Embrittlement ---------------.-.--. - -.- - - ---- 4-9 4.2.1 Upper Shelf Energy---------------------.-.-.-- - -.- - - ---- 4-9 4.2.2 Pressurized Thermal Shock ----------------.-.-.-- - -.- - - ---- 4-10 4.2.2 Pressure-Temperature (P-T) Limits----------.-.-.-- - -.- - - ---- 4-11 Metal Fatigue .......---.-.-.-.-.------------------- - -.- - - ---- 4-13 4.3 4.3.1 ASME Boiler and Pressure Vessel Code,Section III, Class 1 - - -.- - - ---- 4 -1 3 4.3.2 ANSI B31.1 Piping ----------------.-.--. - -.- - - ---- 4-14 4.3.3 Reactor Vessel Underclad Cracking --.-----.- - -.- - - ---- 4 -15 4.3.4 Accumulator Check Valves ------------.--. - -.- - - ---- 4 -16 4.3.5 Reactor Vessel Nozzle-to-Vessel Weld Defect - - - -.- - - - - - 16 4.3.6 Pressurizer Fracture Mechanics Analysis ----.- - -.- - - - - - 17 4.3.7 Environmentally Assisted Fatigue ---.-----.- - -.- - - ---- 4-17 4.3.7.1 Reactor Vessel Locations --.-----.- - -.- - - ---- 4-20 4.3.7.2 Surge Line Locations -----------.- - -.- - - ---- 4-20 4.3.7.3 USAS B31.1 Locations ---.-----.- - -.- - - ---- 4 -2 2 4.4 Environmental Qualification (EQ) of Electric Equipment - -
- ---- 4-24 4.4.1 Solenoid Operated Valves------------.--. - -.- - - ---- 4-27 4.4.1.1 ASCO Solenoid Valve Model X-HAV210 - - - -- .
- ---- 4-27 4.4.1.2 Valcor Solenoid Operated Valve, Model V526-5440 -.- - - - - - 28 4.4.1.3 Valcor Solenoid Operated Valve, Model V526-6130 -.- - - - - - 28 4.4.1.4 Head Vent Solenoid Operated Valves -.------ ---- .
- ---- 4-29 4.4.2 Motors - 30 4.4.2.1 Westinghouse Containment Recirculation Fan Motor - -.- - - ---- 4-30 4.4.2.2 Limitorque Actuators, Outside Containment ---- - -.- - - ---- 4-31 4.4.2.3 Limitorque Actuators, Inside Containment ----- - -.- - - ---- 4-31 4.4.2.4 Limitorque Actuators, PORV Block Valves ---.- - -.- - - ---- 4-32 4.4.2.5 Westinghouse Safety Injection Pump 1N1C Motor - -.- - - ---- 4-33 4.4.2.6 Westinghouse/Reliance Safety Injection Pump 1B Mo3tor - ------- --- 4-33 4.4.2.7 Westinghouse Containment Spray Pump Motors - - -.- - - - - - 34 4.4.2.8 Westinghouse RHR Pump 1A Motor -.-----.-- - -.- - - - - - 34 4.4.2.9 Westinghouse RHR Pump 1 B Motor ----------------.-.--.- - 4-35 Page-vi
Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information TECHNICAL AND ADMINISTRATIVE INFORMATION Contents 4.4.2.10 Westinghouse Hydrogen Recombiner Blower Motor -- ------------ 4-35 4.4.2.11 Reliance AC Random Wound Motor Model 145TCV -- ------------ 4-36 4.4.3 Electrical Penetration Assemblies --- ------------------------------- 4-36 4.4.3.1 Crouse-Hinds Electrical Penetration Assemblies - --------------- 4-36 4.4.3.2 Westinghouse Electrical Penetration Assembly, Model WX32714 ------ 4-37 4.4.3.3 Westinghouse Modular Electrical Penetration Assemblies - --------- 4-38 4.4.4 Heat Shrink Tubing ---- ---------------------------------------- 4-38 4.4.4.1 Raychem WCSF-N Splice Sleeves for Containment Penetration Fan Cooler Motors ----- ----------------------------------------- 4-38 4.4.4.2 Raychem Nuclear Splice Kits - NMCK, NPKC, NPKP, and NPKS ------ 4-39 4.4.4.3 Raychem WCSF-050-N Shim Stock Cable Sleeves - ------------- 4-40 4.4.4.4 Raychem WCSF-N Cable Sleeves -- ------------------------ 4-40 4.4.4.5 Raychem Nuclear Splice Kits, Model NESK -- ------------------ 4-41 4.4.4.6 Raychem Nuclear Splice Kits, Model NPKV -- ------------------ 4-41 4.4.5 Wire and Cable ----- ------------------------------------------ 4-42 4.4.5.1 Kerite 600Volt HTK Insulated, FR Jacketed Power Cable ---------- 4-42 4.4.5.2 Kerite 600 Volt FR Insulated, FR Jacketed Control Cable - --------- 4-42 4.4.5.3 Conax Electric Conductor Seal Assembly -- ------------------- 4-43 4.4.5.4 Anaconda FR-EP and FR-EP/CPE Cable -- ------------------- 4-44 4.4.5.5 General PVC Insulated and Jacketed Control Cable -- ------------ 4-44 4.4.5.6 PVC Instrument Cable Outside Containment (Rome Cable Corp.) ------ 4-45 4.4.5.7 BIW Tefzel ETFE Coaxial and Triaxial Cable - - - --------- 4-46 4.4.5.8 BIW EPR Insulated and CSPE Jacketed Cable -- --------------- 4-46 4.4.5.9 Raychem Flamtrol Shielded Cable -- ------------------------ 4-47 4.4.5.10 General PVC Insulated and Jacketed Control Cable -- ------------ 4-48 4.4.5.11 Containment Electrical Penetration Pigtail Extension Cables - -------- 4-48 4.4.5.12 Coleman Silicone Rubber Instrument Cable Inside Containment - ------ 4-49 4.4.5.13 Okonite Control Cable --- -------------------------------- 4-49 4.4.5.14 Conax Core Exit Thermocouple Connector/Cable Assemblies -------- 4-50 4.4.5.15 Namco Limit Switch Connector/Cable Assemblies -- ------------- 4-51 4.4.5.16 Brand Rex Electrical Cable --- ---------------------------- 4-51 4.4.6 Electrical Connectors ---- --------------------------------------- 4-52 4.4.6.1 Amphenol Triaxial Cable Connector --- ---------------------- 4-52 4.4.6.2 EGS Quick Disconnect Electrical Connectors, ITT Cannon GB-1 ------ 4-52 Page-vii
Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information TECHNICAL AND ADMINISTRATIVE INFORMATION Contents 4.4.6.3 Quick Disconnect Electrical Connectors, EGS- - -.- - - - - - 53 4.4.6.4 States Terminal Blocks Model M-25012 ----------------.-.-- 53 4.4.6.5 AMP Butt Splices, Models 53549-1 and 53550-1 ---------.-.--. 54 4.4.6.6 AMP Butt Splices, Models 52979 and 52980-----------.-.-.-- 54 4.4.7 Resistance Temperature Detectors ......-.-.-.-.------------------ 55 4.4.7.1 Conax Resistance Temperature Detector Models 7N92-1 0000 and 7A22 10000 55 4.4.7.2 Conax Resistance Temperature Detector Model 7DB9-10000 ----.- 56 4.4.7.3 Pyromation Resistance Temperature Detector------------.--.-- 57 4.4.8 Victoreen High Range Radiation Monitor --------.--. ---------- 57 4.4.9 Rosemount Conduit Seal----------------.-.-.-- ---------- 58 4.4.10 Transamerica Delaval Level Switch ------------.--. ---------- 58 4.4.11 Schaevitz Linear Variable Differential Transformer ----- ---------- 59 4.4.12 Hydrogen Recombiner Exciter, Ignitor, and Thermocouples ---------- 59 4.5 Concrete Containment Tendon Prestress -----------.-.-.-- - -- - - - - - - - - - 4-61
- -.- - - - - - - - - - 4 -6 3 4.6 Containment Liner Plate and Penetration Fatigue ----.-----.-
4.7 Other Plant Specific TLAAs .....---.-.-.--------------- - - --. . --- 4 -6 5 4.7.1 Containment Liner Stress----------------.-.-.-- - - -. - -. -4 -6 5 Containment Tendon Fatigue--------------.-.--. - -.- - - - - - - - - - 4 -6 5 4.7.2 4.7.3 Containment Liner Anchorage Fatigue------------.- - -.- - - - - - - - - 6 6 4.7.4 Containment Tendon Bellows Fatigue---------.-.-- - -.- - - - - - - - - - 4-66
- -.- - - - - - - - - - 4-6 7 4.7.5 Crane Load Cycle Limit------------------.-.--.
RCP Flywheel ------- -----------------.-.--. - -.- - - - - - - - - - 4 -6 7 4.7.6 4.7.7 Thermal Aging of Cast Austenitic Stainless Steel (CASS) - - -.- - - - - - - - - - 4-68 Section 4.0 References .......- -.-.- -.-.------------------- -. . . . . . . . . . ..- 4-70 A1.0 Appendix A - Introduction ---------------------.-.-.-- - - --. . --- A -1 A2.0 Programs that Manage the Effects of Aging---------.-.--. - -.- - - - - - - - - - A -1 A2.1 Aging Management Programs --------------------.-.--. - -.- - - - - - - - - - A -1 A2.1.1 Aboveground Carbon Steel Tanks ------------.--.-- - -.- - - - - - - - - - A -1 A2.1.2 ASME Section XI, Subsections IWB, IWC, & IWD Inservice Inspection - - - - - - A-1 Page-viii
Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information TECHNICAL AND ADMINISTRATIVE INFORMATION Contents A2.1.3 ASME Section XI, Subsections IWE & IWL Inservice Inspection - ------------ A-2 A2.1.4 ASME Section XI, Subsection IWF Inservice Inspection -- ----------------- A-2 A2.1.5 Bolting Integrity ----- ------------------------------------------ A-2 A2.1.6 Boric Acid Corrosion ---- --------------------------------------- A-2 A2.1.7 Buried Piping and Tanks Inspection --- ------------------------------ A-3 A2.1.8 Closed-Cycle (Component) Cooling Water System -- -------------------- A-3 A2.1.9 Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements .........-----...- -.---- ------------------------- A-3 A2.1.10 Fire Protection ---- ------------------------------------------- A-3 A2.1.11 Fire Water System ---- ----------------------------------------- A-4 A2.1.12 Flow-Accelerated Corrosion ---- ---------------------------------- A-4 A2.1.13 Fuel Oil Chemistry ---- ----------------------------------------- A-4 A2.1.14 Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems ----- ----------------------------------------------- A-4 A2.1.15 One-Time Inspection ---- --------------------------------------- A-5 A2.1.16 Open-Cycle Cooling (Service) Water System -- ------------------------ A-5 A2.1.17 Periodic Surveillance and Preventive Maintenance -- -------------------- A-5 A2.1.18 Reactor Vessel Head Penetration Inspection -- ------------------------ A-5 A2.1.19 Reactor Vessel Internals ---- ------------------------------------ A-6 A2.1.20 Reactor Vessel Surveillance --- ----------------------------------- A-6 A2.1.21 Spent Fuel Pool Neutron Absorber Monitoring -- ----------------------- A-6 A2.1.22 Steam Generator Tube Integrity --- -------------------------------- A-6 A2.1.23 Structures Monitoring Program --- --------------------------------- A-6 A2.1.24 Systems Monitoring ---- ---------------------------------------- A-7 A2.1.25 Thimble Tubes Inspection ---- ------------------------------------ A-7 A2.1.26 Water Chemistry Control ---- ------------------------------------ A-7 A3.0 Evaluation of Time-Limited Aging Analyses --- ----------------------------- A-8 A3.1 Reactor Vessel Neutron Embrittlement ---- --------------------------------- A-8 A3.1.1 Upper Shelf Energy ---- ---------------------------------------- A-8 A3.1.2 Pressurized Thermal Shock ---- ---------------------------------- A-9 A3.1.3 Pressure-Temperature Limits ---- --------------------------------- A-9 Page-ix
Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information TECHNICAL AND ADMINISTRATIVE INFORMATION Contents A3.2 Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)------------ A-10 A3.3 Metal Fatigue------------ A-10 A3.3.1 ASME Boiler and Pressure Ve ssel Code,Section III, Class 1 -------- ---- --- A-10 A3.3.2 Reactor Vessel Underclad Craacking- --------- --- ---- --- --- ---- --- --- A-i 1 A3.3.3 ANSI B31.1 Piping ---.- --- .....- - - -- - --- --- - - - A -12 A3.3.4 Accumulator Check Valves A-12 A3.3.5 Environmentally Assisted Fati gue --- -------------------- ------------ A-13 A3.3.6 Pressurizer Surge Line - -
A-13 A3.4 Environmental Qualification of Electric Equipment -------- --- ---- --- --- ---- ----- A-14 A3.5 Concrete Containment Tendon Prestre ss -A-15
. . . -.. -.. -.. -.. . .-. .-. .-. .-. .-. . . .-. .- A -1 5 3.5.1 Containment Tendon Fatigue 3.5.2 Containment Tendon Bellows Fatigue --- -------------- --- --- ---- ----- A-16 A3.6 Containment Liner Plate and Penetrati on Fatigue --- -------------- ---- --- --- --- A-16 3.6.1 Containment Liner Anchorage 'Fatigue --- ----------- ---- --- --- --- ---- A-16 A3.7 Containment Liner Stress ----.- --- .....- - - --- -- - --- - - - A -17 A3.8 Other Time-Limited Aging Analyses -- - .....- - - --- - -- --- - - - A -17 A3.8.1 Crane Load Cycle Limit - - -- - .....- - - --- --- --- - - - A -17 A3.9 Exemptions------------.--. --- .....- - - --- - -- --- - - - A -17 A4.0 TLAA Supporting Activities ---- --- .....- - - --- --- - -- - - - A -18 A4.1 Concrete Containment Tendon Prestre ss --- ------------------ ---------- ----- A-18 A4.2 Environmental Qualification Program --- .....- - - - -- --- --- - - - A -18 A4.3 Fatigue Monitoring Program ---.- --- .....- - - --- --- -- - - - - A -18 B1.0 Appendix B - Introduction ----.-
B1.1 Overview-----------.-.-.-- -- - .....- - - --- --- - -- -- - B -1 B1.2 Operating Experience -.-----.- --- .....- - - -- - --- --- --- B -4 B1i.3 Aging Management Programs - - - --- .....- - - - -- --- --- --- B-5 B1.4 Time-Limited Aging Analyses Support Activities: --- ------------ ---- --- --- ------ B-6 B2.0 Aging Management Programs - - - --- .....- - - -- - --- --- --- B-7 Page-x
Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information TECHNICAL AND ADMINISTRATIVE INFORMATION Contents B2.1 Aging Management Activities ---- --------------------------------------- B-i 1 B2.1.1 Aboveground Carbon Steel Tanks --- ------------------------------- B-i 1 B2.1.2 ASME Section Xl, Subsections IWB, IWC, & IWD Inservice Inspection - -------- B-12 B2.1.3 ASME Section Xl, Subsections IWE & IWL Inservice Inspection - ----------- B-14 B2.1.4 ASME Section Xl, Subsection IWF Inservice Inspection -- ----------------- B-16 B2.1.5 Bolting Integrity ----- ------------------------------------------ B-17 B2.1.6 Boric Acid Corrosion ---- --------------------------------------- B-19 B2.1.7 Buried Piping and Tanks Inspection --- ------------------------------ B-20 B2.1.8 Buried Piping and Tanks Surveillance --- ---------------------------- B-21 B2.1.9 Closed-Cycle (Component) Cooling Water System -- -------------------- B-22 B2.1.10 Compressed Air Monitoring ---- ---------------------------------- B-23 B2.1.11 Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements ----- ------------------------------------------- B-24 B2.1.12 Electrical Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits --- -------------------------------- B-28 B2.1.13 Fire Protection ----- ------------------------------------------ B-29 B2.1.14 Fire Water System ---- ----------------------------------------- B-30 B2.1.15 Flow-Accelerated Corrosion ---- ---------------------------------- B-31 B2.1.16 Fuel Oil Chemistry ---- ----------------------------------------- B-33 B2.1.17 Inaccessible Medium Voltage Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements ---- ----------------------------------- B-35 B2.1.18 Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems ----- ----------------------------------------------- B-35 B2.1.19 Loose Part Monitoring ---- -------------------------------------- B-36 B2.1.20 Neutron Noise Monitoring ---- ------------------------------------ B-37 B2.1.21 One-Time Inspection ---- --------------------------------------- B-38 B2.1.22 Open-Cycle Cooling (Service) Water System -- ------------------------ B-41 B2.1.23 Periodic Surveillance and Preventive Maintenance -- -------------------- B-42 B2.1.24 Protective Coatings Monitoring and Maintenance Program -- --------------- B-46 B2.1.25 Reactor Head Closure Studs ---- --------------------------------- B-51 B2.1.26 Reactor Vessel Head Penetration Inspection -- ------------------------ B-52 B2.1.27 Reactor Vessel Internals ---- ------------------------------------ B-55 Page-xi
Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information TECHNICAL AND ADMINISTRATIVE INFORMATION Contents B-60 B2.1.28 Reactor Vessel Surveillance .....................
B-63 B2.1.29 Selective Leaching of Materials ....................
B-63 B2.1.30 Spent Fuel Pool Neutron Absorber Monitoring------------.---
B-66 B2.1.31 Steam Generator Tube Integrity ....................
B-68 B2.1.32 Structures Monitoring Program ....................
B-69 B2.1.33 Systems Monitoring ...... . ... ... ....... ......
B-72 B2.1.34 Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)
Stainless Steel B2.1.35 Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic (CASS) .. ... .. . .. ... .. .. .. .. . .. - -.. . . -. . -. . -. . -. . . .- . D -10
- - - - -.. - - - -- - B-73 B2.1.36 Thimble Tubes Inspection --------------.--.
- - - - -.. - - - -- - B-76 B2.1.37 Water Chemistry Control--------------.--.-
- - - - -.. - - - -- - B-78 B3.0 Time-Limited Aging Analyses Support Activities -----.-
- - - - -.. - - - - - - B-78 B3.1 Environmental Qualification Program ----------.-.-.--
- - - - -.. - - - -- - B-81 B3.2 Fatigue Monitoring . . . . . . . . . . . . . . . . . . . . . . .
. -. . . . . . . . . . . . .- B -8 6 B3.3 Concrete Containment Tendon Pre-stress -----------.-
- - - - -.. - - - -- - B-89 Appendix B References . . . . . . . . . . . . . . . . . . . .
- - - - -.. - - - -- - C-1 C1.0 Appendix C - Not Used ....................
- - - - -.. - - - -- - D-1 D2.0 Appendix D - Technical Specifications Changes -----.-
- - E-1 E1.0 Appendix E - Environmental Report------------.--.--
Page-xii
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information TECHNICAL AND ADMINISTRATIVE INFORMATION Tables 2-28 Table 2.1-1 System Function Codes .............................................
2-35 Table 2.2-1 Plant Level Scoping Results ..........................................
2-45 Table 2.3.1 -1 Reactor Coolant (Class 1) ............................................
2-47 Table 2.3.1-2 R eactor V essel ....................................................
2-51 Table 2.3.1-3 Reactor Vessel Internals .............................................
2-55 Table 2.3.1-4 P ressurizer ................... ................... ............. ....
2-57 Table 2.3.1-5 Steam G enerators ..................................................
2-63 Table 2.3.1-6 Reactor Coolant (Non-Class 1) ........................................
2-69 Table 2.3.2-1 Safety Injection (S I) .................................................
2-75 Table 2.3.2-2 Containm ent Spray (CS) .............................................
2-81 Table 2.3.2-3 Residual Heat Removal (RHR) ........................................
2-86 Table 2.3.2-4 Containment Hydrogen Detectors and Recombiners .......................
2-90 Table 2.3.2-5 Containment Isolation Components ....................................
2-96 Table 2.3.3-1 Chemical and Volume Control (CVCS) ..................................
2-102 Table 2.3.3-2 Component Cooling W ater (CCW ) .....................................
2-109 Table 2.3.3-3 Spent Fuel Cooling and Fuel Storage ...................................
2-115 Table 2.3.3-4 W aste D isposal ....................................................
2-120 Table 2.3.3-5 Service W ater (SW ) .................................................
2-128 Table 2.3.3-6 Fire Protection (FP) .................................................
2-136 Table 2.3.3-7 Heating Steam System ..............................................
2-139 Table 2.3.3-8 Em ergency Pow er ..................................................
2-148 Table 2.3.3-9 Containm ent Ventilation .............................................
2-157 Table 2.3.3-10 Essential Ventilation Systems .........................................
2-161 Table 2.3.3-11 Cranes, Hoists, and Lifting Devices ....................................
2-163 Table 2.3.3-12 T reated Water .....................................................
2-168 Table 2.3.3-13 Radiation Monitoring ................................................
2-187 Table 2.3.4-1 Main and Auxiliary Steam ............................................
2-193 Table 2.3.4-2 Feedwater and Condensate ..........................................
2-199 Table 2.3.4-3 Auxiliary Feedwater (AFW ) ...........................................
2-204 Table 2.3.4-4 Turbine-Generator and Supporting Systems ..............................
2-211 Table 2.4.1-1 Containm ent Structures ..............................................
2-226 Table 2.4.2-1 A uxiliary Building ...................................................
2-238 Table 2.4.2-2 Interm ediate Building ................................................
2-249 Table 2.4.2-3 T urbine Building ...................................................
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information TECHNICAL AND ADMINISTRATIVE INFORMATION Tables Table 2.4.2-4 D iesel Building ..................................................... 2-257 Table 2.4.2-5 C ontrol Building .................................................... 2-268 Table 2.4.2-6 All Volatile Water Treatment Building ................................... 2-277 Table 2.4.2-7 Screen House Building .............................................. 2-286 Table 2.4.2-8 Standby Auxiliary Feedwater Building ................................... 2-294 Table 2.4.2-9 Service Building .................................................... 2-301 Table 2.4.2-10 C able Tunnel ...................................................... 2-308 Table 2.4.2-11 Essential Yard Structures ............................................ 2-314 Table 2.4.2-12 Component Supports Commodity Group ................................ 2-318 Table 2.5.2-1 120 VAC Vital Instrument Buses ....................................... 2-333 Table 2.5.3-1 125 VDC Pow er .................................................... 2-336 Table 2.5.4-1 4160 VAC Pow er ................................................... 2-339 Table 2.5.5-1 480 VAC Power .................................................... 2-344 Table 2.5.6-1 Control Rod Drive and Nuclear Process Instrumentation .................... 2-349 Table 2.5.7-1 Misc. AC Power and Lighting ......................................... 2-352 Table 2.5.8-1 Offsite Pow er ...................................................... 2-354 Table 2.5.9-1 Reactor Protection .................................................. 2-360 Table 2.5.10-1 Engineered Safety Features Actuation System ............................ 2-363 Table 2.5.11-1 Plant Com m unications .............................................. 2-366 Table 2.5.12-1 Plant Process Computers ............................................ 2-368 Table 2.5.13-1 Plant Security ..................................................... 2-370 Table 2.5.14-1 Seismic and Meteorological .......................................... 2-372 Table 3.1-1 Internal Service Environments ........................................ 3-8 Table 3.1-2 External Service Environments ........................................ 3-11 Table 3.2.0-1 Class 1 Piping and Associated Pressure Boundary Components - WCAP-14575-A Final Safety Evaluation Report Response to Applicant Action Items ................ 3-17 Table 3.2.0-2 Reactor Internals - WCAP-14577, Rev. 1-A, Final Safety Evaluation Report Response to Applicant Action Item s ............................................... 3-25 Table 3.2.0-3 Pressurizers - WCAP-14574-A Final Safety Evaluation Report Response to Applicant A ction Item s ....................................................... 3-30 Table 3.2-1 Reactor Coolant System - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal ..................................... 3-40 Table 3.2-2 Reactor Coolant System - Component Types Subject to Aging Management not Evaluated in NU R EG -1801 ................................................... 3-63 Table 3.3-1 Engineered Safety Features Systems - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal ...................... 3-80 Page-xiv
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information TECHNICAL AND ADMINISTRATIVE INFORMATION Tables Table 3.3-2 Engineered Safety Features Systems - Component Types Subject to Aging Management not Evaluated in NUREG-1801 ........................................ 3-83 Table 3.4-1 Auxiliary Systems - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal ........................................ 3-102 Table 3.4-2 Auxiliary Systems - Component Types Subject to Aging Management not Evaluated in NUR EG -1801 ..................................................... 3-113 Table 3.5-1 Steam and Power Conversion Systems - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal ...................... 3-197 Table 3.5-2 Steam and Power Conversion Systems - Component Types Subject to Aging Management not Evaluated in NUREG-1801 ........................................ 3-200 Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Item s ............................................... 3-218 Table 3.6.0-2 Reactor Coolant System Supports - WCAP-14422, Rev. 2-A, Final Safety Evaluation Report Response to Applicant Action Items .............................. 3-246 Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal ...................... 3-259 Table 3.6-2 Structures and Component Supports - Component Types Subject to Aging Management not Evaluated in NUREG-1801 ........................................ 3-282 Table 3.7-1 Electrical and Instrumentation and Controls Systems - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal ........... 3-303 Table 3.7-2 Electrical and Instrumentation and Controls Systems - Component Types Subject to Aging Management not Evaluated in NUREG-1801 ............................. 3-306 Table 4.1-1 Time-Limited Aging Analysis Categories ................................. 4-7 Table 4.4-1 Environmental Qualification Reanalysis Attributes ......................... 4-26 Table B2.0-1 Correlation Between GALL Programs and Ginna Programs .................. B-7 Page-xv
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information TECHNICAL AND ADMINISTRATIVE INFORMATION Figures Scoping and Screening Process Overview ............................... 2-33 Figure 2.1-1 4-3 Figure 4.1-1 TLAA Methodology ................................................
4-23 Figure 4.3-1 TLAA & GSI-1 90 Environmentally Assisted Fatigue Evaluation Process ........
Page-xvi
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 1.0 ADMINISTRATIVE INFORMATION This application has been prepared to provide the administrative, technical and environmental information required by 10 CFR Part 54 (Reference 1) and 10 CFR Part 51 (Reference 2) to support the renewal of the Operating License for R. E. Ginna Nuclear Power Plant (Ginna Station).
Consistent with the expectations in RIS 2001-05 (Reference 3), one hardcopy of this application is being provided for the public document room, with remaining copies on CD-ROM. Another hard copy is being provided to the NRC License Renewal Project Manager. For the reviewer's convenience, the CD-ROMs also contain copies of the Updated Final Safety Analysis Report (UFSAR) and the license renewal (LR) drawings prepared in support of the license renewal effort.
Hyperlinks to those documents are provided in the application where appropriate. Five sets of hardcopy drawings are also being provided to the NRC under a separate cover letter.
The UFSAR, LR drawings, and other references cited within the application are for information only, and are not incorporated by reference into the LRA.
This section of the application provides the following information:
- 1. Information on the organization of the application (Section 1.1).
- 2. A general plant description (Section 1.2).
- 3. The administrative information required by 10 CFR 54.17 and 54.19 (Section 1.3).
- 4. Summary of abbreviations and passive function code definitions (Section 1.5).
- 5. A distribution list for written communications related to the application (Section 1.6).
Page 1-1
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 1.1 Application Format and Content The following discussion describes the content of the Ginna Station License Renewal Application.
Section 1 provides the administrative information required by Part 54 of Title 10 of the Code of Federal Regulations, Sections 17 and 19 (10 CFR 54.17 and 10 CFR 54.19).
Section 2 provides the scoping and screening methodology. Section 2 also describes and justifies the methodology used to determine the systems, structures, and components within the scope of license renewal and the structures and components subject to an aging management review. The system groupings in Sections 2 and 3 are organized to be consistent with NUREG-1800. Table 2.2-1, Plant Level Scoping Results, provides listings of the plant mechanical systems, structures, and electrical/instrumentation and controls (I&C) systems, and identifies those plant systems and structures that are and are not within the scope of license renewal. Sections 2.3, 2.4 and 2.5 provide a description of systems, their intended functions, and for information only, cross references to UFSAR sections and license renewal drawings. Each system subsection has a table listing component groups subject to an Aging Management Review (AMR), their passive intended function, and one or more hyperlinked cross references to the Section 3 table line items providing AMR information. The drawings and UFSAR are provided as a separate attachment for use as review tools.
Section 3 describes the results of the aging management reviews for the components and structures requiring aging management reviews. Section 3 identifies the components and structures subject to aging management review including a comparison to the structures and components identified in the U. S. Nuclear Regulatory Commission's (NRC) "Generic Aging Lessons Learned (GALL) Report," NUREG-1801, which are combined into Standard Review Plan System groups. Those Ginna Station component groups with aging effects, and aging management programs selected to manage those effects which are consistent with the assumptions made in the GALL Report, are contained in a set of tables that are identical to those in GALL Volume I and the Standard Review Plan. A second set of tables in each Section 3 system grouping provides aging management information including Component group, material, environment, aging effect and selected aging management programs for those component groups where one or more of the column details were not consistent with that assumed in the GALL Report. All the Section 3 tables have a final column that contains additional explanatory information specific to that line item. In addition these tables have hyperlinked cross references to the aging management details in Appendix B.
Section 4 includes a list of time-limited aging analyses (TLAAs), as defined by 10 CFR 54.3. It includes the identification of the component or subject, and an explanation of the time dependent aspects of the calculation or analysis. Section 4 demonstrates that the analyses Page 1-2
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information remain valid for the period of extended operation, the analyses have been projected to the end of the period of extended operation, or the effects of aging on the intended function(s) will be adequately managed for the period of extended operation. Section 4 also states that no 10 CFR 50.12 exemption involving a time-limited aging analysis as defined in 10 CFR 54.3 is required during the period of extended operation. There are some issues discussed in Section 4 that are not considered TLAAs. These are included to provide information on some issues that have been routinely discussed as part of Section 4 in past License Renewal Applications.
Appendix A, Updated Final Safety Analysis Report Supplements, contains a summary description of the programs for managing the effects of aging for the period of extended operation. A summary description of the evaluation of time-limited aging analyses for the period of extended operation is also included.
Appendix B, Aging Management Programs, describes the aging management programs and activities and demonstrates that the aging effects on the components and structures within the scope of the License Renewal Rule will be managed such that they will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation. Where the Ginna Station programs are consistent with corresponding programs in the GALL Report, the appropriate GALL program is referenced.
Appendix C is not used for this application.
Appendix D, Technical Specification Changes, concludes that no technical specification changes are necessary to manage the effects of aging during the period of extended operation.
Appendix E, Environmental Report, contains an environmental report analyzing the potential environmental impacts of license renewal, as provided for in NRC regulations 10 CFR 51.53(c) and 10 CFR 54.23. The NRC requires license renewal applicants to provide the NRC with input, in the form of an environmental report, that the NRC will use to meet NEPA requirements as they apply to license renewal [10 CFR 51.53(c)].
The NRC has determined that nuclear power plant license renewal decisions are major federal actions requiring preparation of an environmental impact statement [10 CFR 51.20(a)(2) and 51.95(c)]. In an effort to streamline the license renewal environmental review, the NRC conducted a generic analysis and published the results in NUREG-1437, Generic Environmental Impact Statement for the License Renewal of Nuclear Power Plants (GELS). To fulfill NEPA requirements, the NRC is required to publish site-specific analyses in the form of a supplemental environmental impact statement to the GELS.
The information in Section 2, Section 3, and Appendix B fulfills the requirements in 10 CFR 54.21 (a). Section 1.4 discusses how the requirements of 10 CFR 54.21 (b) will be met. The information in Section 4 fulfills the requirements in 10 CFR 54.21(c). The information in Page 1-3
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Appendix A and Appendix D fulfills the requirements in 10 CFR 54.21 (d) and 10 CFR 54.22, respectively. The Environmental Report, as required by 10 CFR 54.23, is provided with the Ginna Station License Renewal Application as Appendix E.
1.2 Plant Description Ginna Station is a two-loop pressurized water reactor, rated at 490 MWe, located in Wayne County, near Rochester, New York. Ginna Station is located on the south shore of Lake Ontario, which is the source of circulating water and is the ultimate heat sink. The site, including the switchyard, contains 488 acres. The turbine and condenser system as well as the nuclear steam supply system were designed and supplied by Westinghouse. The remainder of the plant was designed by either Rochester Gas and Electric Corporation or Gilbert Associates, Incorporated. The replacement steam generators were designed and supplied by Babcock and Wilcox International.
1.3 Information Required by 10 CFR 54.17 and 10 CFR 54.19 1.3.1 Name of Applicant Rochester Gas and Electric Corporation (RG&E)
RG&E is currently the sole owner and operator of Ginna Station. RGS Energy Group, Inc.
(RGS), the parent company of RG&E, is a wholly-owned subsidiary of Energy East Corporation (Energy East).
1.3.2 Address of Applicant Rochester Gas and Electric Corporation 89 East Avenue Rochester, NY 14649 RGS Energy Group, Inc.
89 East Avenue Rochester, NY 14649 Energy East Corporation Commerce Plaza Albany, NY 12260 Page 1-4
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 1.3.3 Description of Business or Occupation of Applicant RG&E is a New York corporation engaged principally in the generation of electricity and the purchase, transmission, distribution and sale of electric power and natural gas in Western New York State. RG&E operates under the general regulatory supervision of the New York State Public Service Commission (NYPSC) and, for its wholesale electricity sales and related interstate activities, RG&E is subject to regulation by the Federal Energy Regulatory Commission (FERC). With respect to its interests in Ginna Station, RG&E recovers its cost of generating electricity through rates subject to the regulatory authority of the NYPSC and FERC.
RG&E is a wholly-owned indirect subsidiary of Energy East, a registered public utility holding company under the Public Utility Holding Company Act of 1935 (PUHCA). Energy East has a number of wholly-owned public utility subsidiaries in the northeastern United States, including New York State Electric & Gas Corporation (NYSEG).
Energy East acquired 100% of the common stock of RGS, which is currently a direct wholly-owned subsidiary of Energy East. RG&E is a direct wholly-owned subsidiary of RGS.
Energy East has transferred all of NYSEG's common stock to RGS, so that RG&E and NYSEG can be operated under a combined management structure.
NRC approval of the indirect transfer, as a result of the acquisition of RGS by Energy East, of RG&E's NRC-issued licenses was required. Accordingly, RG&E submitted an application pursuant to Section 50.80. This application was approved on December 10, 2001. The application addressed the issues central to the NRC's review of an indirect license transfer request, including the purpose and nature of the proposed transaction, as well as the impact of the proposed transaction on RG&E's existing financial and technical qualifications, its ability to provide decommissioning funding assurance, and the absence of foreign control or domination over the parent company.
As a result of the recent acquisition of RGS by Energy East, RG&E is now an indirect wholly-owned subsidiary of Energy East. However, this recent change in its ownership has not altered RG&E's status as a distinct corporate entity, nor its status as a regulated utility.
RG&E continues to own the assets that it owned prior to the RGS acquisition and continues to engage in the generation of electricity and the purchase, transmission, distribution and sale of electric power and natural gas in its existing service territories. RG&E continues to maintain sole ownership of Ginna Station, and operates the station as an NRC licensee.
RG&E continues to recover its costs associated with generation electricity at this facility through rates subject to the regulatory authority of the NYPSC and the FERC.
1.3.4 Organization and Management of Applicant Rochester Gas and Electric Corporation Rochester Gas and Electric Corporation is not owned, controlled or dominated by an alien, a foreign corporation, or a foreign government. All officers and directors are citizens of the Page 1-5
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information United States of America. The names and addresses of the directors and principal officers are provided below:
Directors Name Address Wesley W. von Schack Commerce Plaza Albany, NY 12260 Kenneth M. Jasinski Commerce Plaza Albany, NY 12260 Paul C. Wilkens 89 East Avenue Rochester, NY 14649 Princioal Officers Name Address Paul C. Wilkens 89 East Avenue President Rochester, NY 14649 Louis L. Bellina 89 East Avenue Vice-President, Customer Relations Rochester, NY 14649 David J. Irish 89 East Avenue Vice-President, Fossil/Hydro Operations Rochester, NY 14649 Mark Keogh 89 East Avenue Vice-President, Treasurer and Secretary Rochester, NY 14649 Robert C. Mecredy 89 East Avenue Vice-President, Nuclear Operations Rochester, NY 14649 Clifton B. Olson 89 East Avenue Vice-President, Energy Supply Rochester, NY 14649 Jessica S. Raines 89 East Avenue Vice-President, Support Services Rochester, NY 14649 Paul G. Ruganis 89 East Avenue Vice-President, Information Services Rochester, NY 14649 Page 1-6
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information William L. Thomas 89 East Avenue Vice-President, Human Resource Rochester, NY 14649 Services Michael B. Whitcraft 89 East Avenue Vice-President, Energy Delivery Rochester, NY 14649 Joseph A. Widay 89 East Avenue Vice-President and Plant Manager, Rochester, NY 14649 Ginna Station Joseph J. Syta 89 East Avenue Controller Rochester, NY 14649 Kathleen C. Spellane 89 East Avenue Assistant Treasurer Rochester, NY 14649 RGS Energy Group, Inc.
RGS Energy Group, Inc. is not owned, controlled or dominated by an alien, a foreign corporation, or a foreign government. All officers and directors are citizens of the United States of America. The names and addresses of the directors and principal officers are provided below:
Directors Name Address Wesley W. von Schack Commerce Plaza Albany, NY 12260 Kenneth M. Jasinski Commerce Plaza Albany, NY 12260 Paul C. Wilkens 89 East Avenue Rochester, NY 14649 Principal Officers Name Address Wesley W. von Schack Commerce Plaza Chairman, President, and Chief Albany, NY 12260 Executive Officer Page 1-7
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Kenneth M. Jasinski Commerce Plaza Executive Vice President and Chief Albany, NY 12260 Financial Officer Robert D. Kump Commerce Plaza Vice President and Secretary Albany, NY 12260 Energy East Corporation Energy East Corporation is not owned, controlled or dominated by an alien, a foreign corporation, or a foreign government. All officers and directors are citizens of the United States of America. The names and addresses of the directors and principal officers are provided below:
Directors Name Address Richard Aurelio Commerce Plaza Albany, NY 12260 James A. Carrigg Commerce Plaza Albany, NY 12260 Joseph J. Castiglia Commerce Plaza Albany, NY 12260 Lois B. DeFleur Commerce Plaza Albany, NY 12260 G. Jean Howard Commerce Plaza Albany, NY 12260 David M. Jagger Commerce Plaza Albany, NY 12260 John M. Keeler Commerce Plaza Albany, NY 12260 Ben E. Lynch Commerce Plaza Albany, NY 12260 Page 1-8
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Peter J. Moynihan Commerce Plaza Albany, NY 12260 Walter G. Rich Commerce Plaza Albany, NY 12260 Wesley W. von Schack Commerce Plaza Albany, NY 12260 Principal Officers Name Address Wesley W. von Schack Commerce Plaza Chairman, President and Chief Albany, NY 12260 Executive Officer Kenneth M. Jasinski Commerce Plaza Executive Vice-President and Chief Albany, NY 12260 Financial Officer Robert D. Kump Commerce Plaza Vice-President, Treasurer and Secretary Albany, NY 12260 Robert E. Rude Commerce Plaza Vice-President and Controller Albany, NY 12260 1.3.5 Class of License, Use of Facility, and Period of Time for which the License is Sought RG&E requests renewal of the Class 103 operating license for Ginna Station (License No.
DPR-18) for a period of 20 years beyond the expiration date of the current license on September 18, 2009.
RG&E also requests renewal of the source, special nuclear material, and by-product license that is included within the operating license and that was issued pursuant to 10 CFR Parts 30, 40, and 70.
1.3.6 Earliest and Latest Dates for Alterations, if Proposed RG&E does not propose to alter the station in connection with this application. The current licensing basis (CLB) will be continued and maintained throughout the period of extended operation.
Page 1-9
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 1.3.7 Listing of Regulatory Agencies Having Jurisdiction and News Publications The Federal Energy Regulatory Commission (FERC) and the New York State Public Service Commission (NYPSC) are the principal regulators of the company's electric operations.
The Honorable David P. Boergers Secretary Federal Energy Regulatory Commission 888 First Street, NE, Room 1A Washington, DC 20426 Janet H. Deixler Secretary to the Commission New York State Public Service Commission New York State Department of Public Service 3 Empire State Plaza Albany, NY 12223-1350 The area news publication and its associated address is provided below Democrat and Chronicle 55 Exchange Blvd Rochester, NY 14614 1.3.8 Conforming Changes to Standard Indemnity Agreement 10 CFR 54.19(b) requires that license renewal applications include, "...conforming changes to the standard indemnity agreement, 10 CFR 140.92, Appendix B, to account for the expiration term of the proposed renewed license." The current indemnity agreement for the unit does not contain a specific expiration term for the operating license. Therefore, conforming changes to account for the expiration term of the proposed renewed license are not necessary, unless the license number is changed upon issuance of the renewed license.
1.3.9 Restricted Data Agreement This application does not contain restricted data or other national defense information, nor is it expected that subsequent amendments to the license application will contain such information. However, pursuant to 10 CFR 54.17(g) and 10 CFR 50.37, RG&E, as a part of the application for a renewed operating license, hereby agrees that it will not permit any individual to have access to or any facility to possess Restricted Data or classified National Page 1-10
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Security Information until the individual and/or facility has been approved for such access under the provisions of 10 CFR Parts 25 and/or 95.
1.4 Current Licensing Basis Changes During NRC Review Each year, following the submittal of the Ginna Station License Renewal Application and at least three months before the scheduled completion of the NRC review, Ginna Station will submit amendments to the application pursuant to 10 CFR 54.21(b). These revisions will identify any changes to the current licensing basis that materially affect the contents of the License Renewal Application, including the UFSAR supplements and any other aspects of the application.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 1.5 Abbreviations and Passive Function Code Definitions 1.5.1 Abbreviations This section contains the abbreviations that pertain to the administrative and technical information within the license renewal application. The abbreviations that pertain to the environmental information are included in the front of Appendix E (Environmental Report).
Abbreviation Definition AC Alternating Current ABVS Auxiliary Building Ventilation System ACI American Concrete Institute AEC Atomic Energy Commission AFW Auxiliary Feedwater AISC American Institute of Steel Construction AMP Aging Management Program AMR Aging Management Review AMSAC ATWS Mitigation System Actuation Circuitry ANSI American National Standards Institute AR ACTION Report ARV Atmospheric Relief Valve ASCO Automatic Switch Company ASME American Society of Mechanical Engineers ASTM American Society for Testing and Materials ATWS Anticipated Transients Without a SCRAM AVT All-Volatile-Treatment BADGER Boron Areal Density Gauge for Evaluating Racks BAST Boric Acid Storage Tank BF3 Boron Trifluoride BSS Borated Stainless Steel BTP Branch Technical Position B&W Babcock and Wilcox Page 1-12
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Abbreviation Definition BWR Boiling Water Reactor CAR Corrective Action Report CASS Cast Austenitic Stainless Steel CCW Component Cooling Water CD-ROM Compact Disk-Read only Memory CFR Code of Federal Regulations CLB Current Licensing Basis CLOC Closed Loop Outside Containment CMIS Configuration Management Information System CREATS Control Room Emergency Air Treatment System CRDM Control Rod Drive Mechanism CRFC Containment Recirculation Fan Cooling CRT Cathode Ray Tube CS Containment Spray or Carbon Steel CUF Cumulative Usage Factor CVCS Chemical and Volume Control System CW Circulating Water DAM Data Acquisition Modules DBA Design Basis Accident DBD Design Basis Document DBE Design Basis Event DC Direct Current DMIMS Digital Metal Impact Monitoring System DOJ U.S. Department of Justice ECCS Emergency Core Cooling System EDG Emergency Diesel Generator EFPY Effective Full Power Year Page 1-13
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Abbreviation Definition EMPA Swiss Federal Testing Station Energy East Energy East Corporation EPM-FPPR Fire Protection Program Report EPR Ethylene Propylene Rubber EPRI Electric Power Research Institute EQ Environmental Qualification ESF Engineered Safety Features ESFAS Engineered Safety Features Actuation System FAC Flow-Accelerated Corrosion FCC Federal Communications Commission Fen Environmental Fatigue Multiplier FERC Federal Energy Regulatory Commission FME Foreign Material Exclusion FP Fire Protection FRP Fiberglass Reinforced Plastic FSER Final Safety Evaluation Report GAI Gilbert Associates, Inc.
GALL Generic Aging Lessons Learned GDC General Design Criterion GElS Generic Environmental Impact Statement GL Generic Letter GSI Generic Safety Issue GTR Generic Technical Report HAZ Heat-Affected Zone HELB High Energy Line Break HEPA High Efficiency Particulate Air HIC High Integrity Container Page 1-14
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Abbreviation Definition HSLAS High Strength Low Alloy Steel HTK High Temperature Kerite HUT Holdup Tank HVAC Heating, Ventilation, and Air Conditioning I&C Instrumentation and Controls IASCC Irradiation-Assisted Stress Corrosion Cracking ID Inner Diameter IDR Identified Deficiency Report IE Inspection and Enforcement IGA Intergranular Attack IGSCC Intergranular Stress Corrosion Cracking IN Information Notice INEL Idaho National Engineering Laboratories INPO Institute of Nuclear Power Operations IPA Integrated Plant Assessment IR Insulation Resistance ISI Inservice Inspection ITG Issues Task Group LBB Leak-Before-Break LER Abnormal Occurrence and Licensee Event Report LOCA Loss-of-Coolant Accident LR License Renewal LRA License Renewal Application LRE License Renewal Engineer LTOPS Low Temperature Overpressure Protection System LVDT Linear Variable Differential Transformer MDAFW Motor-Driven Auxiliary Feedwater Page 1-15
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Abbreviation Definition MIC Microbiologically Induced Corrosion MOV Motor Operated Valve MRP Materials Reliability Program MRPI Microprocessor Rod Position Indication MRV Minimum Required Value MSIV Main Steam Isolation Valve MSSV Main Steam Safety Valve MT Magnetic Particle Test MWe Megawatt-electric NADP National Atmospheric Deposition Program NaOH Sodium Hydroxide NCR Non-Conformance Report NDE Non-destructive Examination NEI Nuclear Energy Institute NFPA National Fire Protection Association NNS Non-Nuclear Safety NPS Nominal Pipe Size NRC Nuclear Regulatory Commission NSR Non-Safety Related NSAC Nuclear Safety Analysis Center NSSS Nuclear Steam Supply System NTN National Trends Network NYPSC New York State Public Service Commission NYSEG New York State Electric and Gas ODSCC Outside Diameter Stress Corrosion Cracking PE Polyethylene PLL Predicted Lower Limit Page 1-16
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Abbreviation Definition PORV Power Operated Relief Valve PPCS Plant Process Computer System PSPM Periodic Surveillance and Preventive Maintenance Program PSS Plant Sampling System PSSL Plant Systems and Structures List PTLR Pressure-Temperature Limit Report PTS Pressurized Thermal Shock P-T Pressure-Temperature PUHCA Public Utility Holding Company Act of 1935 PVC Polyvinyl Chloride PWR Pressurized Water Reactor PWSCC Primary Water Stress Corrosion Cracking QA Quality Assurance RAI Request for Additional Information RCCA Rod Cluster Control Assembly RCP Reactor Coolant Pump RCPB Reactor Coolant Pressure Boundary RCS Reactor Coolant System RG Regulatory Guide RGS RGS Energy Group, Inc.
RG&E Rochester Gas and Electric Corporation RHR Residual Heat Removal RIO Refueling, Inspection and Overhaul Report RMW Reactor Makeup Water RPV Reactor Pressure Vessel RTD Resistance Temperature Detector Page 1-17
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Abbreviation Definition RTPTS Reference temperature for pressurized thermal shock RTNDT Reference nil ductility transition temperature ARTNDT Radiation-induced shift in the reference nil ductility transition temperature RTS Reactor Trip System RV Reactor Vessel RVI Reactor Vessel Internals RWST Refueling Water Storage Tank SAFW Standby Auxiliary Feedwater SAS Safety Assessment System SBO Station Blackout SC-1 Safety Class 1 SC-2 Safety Class 2 SC-3 Safety Class 3 SCBA Self-Contained Breathing Apparatus SCC Stress Corrosion Cracking SEC Securities and Exchange Commission SEP Systematic Evaluation Program SER Safety Evaluation Report SFP Spent Fuel Pool SFR System Function Report SG Steam Generator SI Safety Injection SOER INPO Significant Operating Event Report SOV Solenoid Operated Valve SPING System-Level Particulate, Iodine, and Nobel Gas Monitors Page 1-18
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Abbreviation Definition SQUG Seismic Qualification Utility Group SR Safety Related SRP Standard Review Plan SS Safety Significant or Stainless Steel SSC Structure System and Component SW Service Water SwRl Southwest Research Institute SWSROP Service Water System Reliability Optimization Program TDAFW Turbine-Driven Auxiliary Feedwater TLAA Time-Limited Aging Analyses TRM Technical Requirements Manual TSC Technical Support Center TS Technical Specification UFSAR Updated Final Safety Analysis Report UHS Ultimate Heat Sink USE Upper Shelf Energy USI Unresolved Safety Issue UT Ultrasonic Testing UV Ultraviolet VCT Volume Control Tank VT Visual Test WOG Westinghouse Owner's Group 1.5.2 Passive Function Code Definitions This section contains the meanings for the abbreviations used in the Screening results tables to represent the passive functions for components, subcomponents, and structural members. Passive functions are the specific intended functions performed by in-scope passive components in support of system or structure intended functions. Passive Page 1-19
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information components are components that perform an intended function without moving parts or without a change in configuration or properties.
CODE DEFINITION CORE SUPPORT CORE SUPPORT DELIVER VOLTAGE, TO PROVIDE ELECTRICAL CONNECTIONS TO CURRENT, OR SIGNALS SPECIFIED SECTIONS OF AN ELECTRICAL CIRCUIT TO DELIVER VOLTAGE, CURRENT, OR SIGNALS DIRECT FLOW STR-PROVIDE SPRAY SHIELD OR CURBS FOR DIRECTING FLOW (E.G. SAFETY INJECTION FLOW TO CONTAINMENT SUMP)
DIRECT GASEOUS STR-PROVIDE PATH FOR RELEASE OF FILTERED DISCHARGE AND UNFILTERED GASEOUS DISCHARGE FASTENING PROVIDE CONNECTION FASTENING FILTER STR-PROVIDE FILTRATION FIRE BARRIER STR-PROVIDE RATED FIRE BARRIER TO CONFINE OR RETARD A FIRE FROM SPREADING TO OR FROM ADJACENT AREAS OF THE PLANT FLAME SUPPRESSION FLAME SUPPRESSION FLOOD BARRIER STR-PROVIDE FLOOD PROTECTION BARRIER (INTERNAL AND EXTERNAL FLOODING EVENT)
FLOW CONTROL FLOW CONTROL FLOW DISTRIBUTION FLOW DISTRIBUTION GUIDE AND SUPPORT GUIDE AND SUPPORT INSTRUMENTATION INSTRUMENTATION GUIDE AND SUPPORT GUIDE AND SUPPORT RCCA'S RCCA'S GUIDE AND SUPPORT GUIDE AND SUPPORT THERMOCOUPLES THERMOCOUPLES HEAT SINK STR-PROVIDE HEAT SINK DURING SBO OR DESIGN BASIS ACCIDENTS Page 1-20
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information CODE DEFINITION HEAT TRANSFER HEAT TRANSFER HELB SHIELDING STR-PROVIDE SHIELDING AGAINST HIGH ENERGY LINE BREAKS HOUSE, PROTECT HOUSE, PROTECT EQUIPMENT EQUIPMENT INSULATE AND INSULATE AND SUPPORT ELECTRICAL SUPPORT CONDUCTOR JOINT INTEGRITY MECHANICAL CLOSURE INTEGRITY LIFTING EQUIPMENT LIFTING EQUIPMENT MAINTAIN ELEC SYSTEM MAINTAIN ELEC SYSTEM INTEGRITY INTEGRITY MISSILE BARRIER STR-PROVIDE MISSILE BARRIER (INTERNALLY OR EXTERNALLY GENERATED)
MOUNT & SUPPORT MOUNT & SUPPORT ELEC COMP & INST ELEC COMP & INST NO NEI 95-10 PASSIVE NO NEI 95-10 PASSIVE FUNCTION FUNCTION NOT EVALUATED INTENDED FUNCTION NOT REQUIRED TO BE EVALUATED PIPE WHIP RESTRAINT STR-PROVIDE PIPE WHIP RESTRAINT PRESSURE BOUNDARY STR-PROVIDE PRESSURE BOUNDARY OR ESSENTIALLY LEAK TIGHT BARRIER TO PROTECT PUBLIC HEALTH AND SAFETY IN THE EVENT OF ANY POSTULATED DESIGN BASIS EVENT PRESSURE BOUNDARY PRESSURE BOUNDARY PROVIDE ELECTRICAL PROVIDE ELECTRICAL ISOLATION ISOLATION PROVIDE FILTRATION PROVIDE FILTRATION PROVIDE FIRE BARRIER PROVIDE FIRE BARRIER PROVIDE FLOW PROVIDE FLOW Page Page 1-21
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information CODE DEFINITION PROVIDE ISOLATION PROVIDE ISOLATION BARRIER BARRIER PROVIDE MIXED FLOW PROVIDE MIXED FLOW PROVIDE RADIATION PROVIDE RADIATION SHIELD SHIELD PROVIDE STRUCTURAL PROVIDE STRUCTURAL SUPPORT SUPPORT RADIATION SHIELDING STR-PROVIDE SHIELDING AGAINST RADIATION AND/OR HEAT RESERVOIR RESERVOIR RESTRICTS FLOW RESTRICTS FLOW RX CLNT PRESS RX CLNT PRESS BOUNDARY BOUNDARY SHELTER SR STR-PROVIDE SHELTER/PROTECTION TO SAFETY-RELATED COMPONENTS SHIELD VESSEL SHIELD VESSEL STRUCTURAL SUPPORT STRUCTURAL SUPPORT SUPPORT IN-CORE SUPPORT IN-CORE INSTRUMENTATION INSTRUMENTATION SUPPORT NSR STR-PROVIDE STRUCTURAL SUPPORT TO NONSAFETY-RELATED COMPONENTS WHOSE FAILURE COULD PREVENT SATISFACTORY ACCOMPLISHMENT OF ANY OF THE REQUIRED SAFETY RELATED FUNCTIONS SUPPORT PIPE, CABLE, SUPPORT PIPE, CABLE, DUCT DUCT SUPPORT RV SUPPORT RV INTERNALS INTERNALS SUPPORT SR STR-PROVIDE STRUCTURAL AND/OR FUNCTIONAL SUPPORT TO SAFETY RELATED EQUIPMENT Page 1-22
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information CODE DEFINITION SUPPORT THIMBLE SUPPORT THIMBLE TUBES TUBES THROTTLING THROTTLING WATER SOURCE STR-PROVIDE SOURCE OF COOLING WATER FOR PLANT SHUTDOWN Page 1-23
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 1.6 Communications Written communications on this application should be directed to:
Mr. Robert C. Mecredy Vice-President, Nuclear Operations R. E. Ginna Nuclear Power Plant 1503 Lake Road Ontario, NY 14519 With copies to:
Mr. George J. Wrobel Manager, Nuclear Safety and Licensing R. E. Ginna Nuclear Power Plant 1503 Lake Road Ontario, NY 14519 And Mr. David F. Wilson R. E. Ginna Nuclear Power Plant 1503 Lake Road Ontario, NY 14519 Page 1-24
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Section 1.0 References
- 1. 10 CFR 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants, U.S.
Nuclear Regulatory Commission.
- 2. 10 CFR 51, Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions, U.S. Nuclear Regulatory Commission.
- 3. RIS 2001-05, NRC Regulatory Issue Summary 2001-05, Guidance on Submitting Documents to the NRC By Electronic Information Exchange or on CD-ROM
- 4. NEI 95-10, Industry Guideline for Implementing the Requirements of 10 CFR Part 54 - The License Renewal Rule, Rev. 3, Nuclear Energy Institute, March 2001.
- 5. Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, NUREG-1 800, U.S. Nuclear Regulatory Commission, July 2001.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 2.0 SCOPING AND SCREENING METHODOLOGY FOR IDENTIFYING STRUCTURES AND COMPONENTS SUBJECT TO AGING MANAGEMENT REVIEW, AND IMPLEMENTATION RESULTS 2.1 Scoping and Screening Methodology 2.1.1 Introduction The initial step in scoping is defining the entire plant in terms of major systems and structures. All of these systems and structures are evaluated against the scoping criteria in 10 CFR 54.4 (a)(1), (2), and (3), to determine if they perform, support or could adversely impact a critical safety function for responding to a design basis accident event, or perform or support a specific requirement of one of five regulated events.
This step is accomplished using the UFSAR, Technical Specifications, licensing correspondence files, Design Basis Documents (DBDs), controlled drawings, the Q-list and the Configuration Management Information System (CMIS), a controlled data base which stores equipment and Licensing Basis information.
During the scoping process, even if only a portion of a system or structure meets the scoping criteria of 10 CFR 54.4, the system or structure is identified as in the scope of License Renewal for subsequent screening with exceptions as indicated in Section 2.1.3.
The scoping and screening process provides a system description and identifies specific system/structure intended functions. System scoping establishes all the components and structural parts of which the system or structure is composed, defines component internal and external environments, and then specifies which components and structural parts support the specific system/structure intended functions (i.e., are within the License Renewal evaluation boundary). As a result, not all of the components or structural parts that make up in-scope systems and structures may be within the scope of License Renewal, since those components do not support an intended function. The screening process identifies those components subject to aging management review. By identifying the active or passive nature of the SSC and whether the SSC is long-lived, only the appropriate components are forwarded to aging management review. The screening process also defines the License Renewal component passive intended functions (e.g. pressure boundary, restrict flow, etc.).
2.1.2 Plant Level Scoping 10 CFR 54 provides specific criteria for determining which systems, structures, and components should be reviewed and evaluated for inclusion in the scope of License Renewal. Specifically, Section 54.4 of the rule states that:
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information (a) Plant systems, structures, and components within the scope of this part are:
(1) Safety related systems, structures, and components which are those relied upon to remain functional during and following design basis events (as defined in 10 CFR 50.49(b)(1)) to ensure the following functions:
(i) The integrity of the reactor coolant pressure boundary; (ii) The capability to shut down the reactor and maintain it in a safe shutdown condition; or (iii) The capability to prevent or mitigate the consequences of accidents that could result in potential off-site exposures comparable to those referred to in 10 CFR 50.34(a)(1), 10 CFR 50.67(b)(2), 10 CFR 100.11, as applicable.
(2) All non-safety-related systems, structures, and components whose failure could prevent satisfactory accomplishment of any of the functions identified in paragraphs (a)(1)(1), (ii), or (iii) above.
(3) All systems, structures, and components relied on in safety analyses or plant evaluations to perform a function that demonstrates compliance with the NRC's regulations for fire protection (10 CFR 50.48), environmental qualification (10 CFR 50.49), pressurized thermal shock (10 CFR 50.61),
anticipated transients without scram (10 CFR 50.62), and station blackout (10 CFR 50.63).
(b) The intended functions that these systems, structures, and components must be shown to fulfill in 54.21 are those functions that are the bases for including them within the scope of License Renewal as specified in paragraphs (a)(1) - (3) of this section.
All plant systems and structures that comprise Ginna Station are reviewed and evaluated against the criteria outlined above to determine which ones meet the requirements for inclusion in the scope of License Renewal. The scoping methodology used by RG&E is consistent with the guidance provided by the NRC in NUREG-1 800 and NEI 95-10. Existing plant documentation is used for this review including the UFSAR, Technical Specifications, and the licensing correspondence files that collectively form the Ginna Current Licensing Basis, controlled drawings, and the Configuration Management Information System (CMIS) database. Once the systems and structures have been identified for inclusion in the scope, they will be further evaluated with the next step in the IPA process - screening.
2.1.3 System and Structure Function Determination The plant 0-list documents Structure System and Component (SSC) Safety Functions in a very structured and comprehensive way. The License Renewal Scoping and Screening Page 2-2
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Process initially uses the rule based classification information and process as a direct input.
Approved plant procedures govern the process for safety classifications while the CMIS database stores the resultant process output.
Plant structures, systems and components are sorted and tracked within CMIS using a system identifier known as the Plant Systems and Structures List (PSSL) number. This detailed numbering scheme supports plant needs with respect to the maintenance rule and maintenance work and offers a good starting point with respect to identifying license renewal system functional boundaries. License Renewal (LR) systems are based on the PSSL and account for and contain all of the PSSL systems, but do so in a manner that is much more consistent with the broader system descriptions in the UFSAR.
SSC safety classifications are functionally based and are valid no matter what system naming or sorting scheme is used. Therefore, describing unique system boundaries for License Renewal has no impact on the ability to determine if an SSC performs a License Renewal intended function.
The SSC functions detailed in the systems scoping reports are derived from the plant's licensing basis. System (and Structure) Function Codes have been assigned and can be related to the criterion in the American National Standard Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants (ANSI/ANS-51.1-1983) should further clarification be required. Table 2.1-1 details the system function codes used.
Systems associated with function codes A through R are systems that perform nuclear safety functions (10 CFR 54.4(a)(1) Criterion 1 functions). Systems associated with function code Y contain components whose failure can affect a nuclear safety function (10 CFR 54.4(a)(2) Criterion 2 functions). Systems associated with function codes Z1 through Z5 contain components that are associated with the License Renewal regulated event set (10 CFR 54.4(a)(3) Criterion 3 functions).
The safety classification process also evaluates how a system component supports the system's safety function. This evaluation accounts for both active and passive design functions (e.g. pressure boundary, a passive function and allow flow, an active function).
The classification evaluation is documented in CMIS in the form of numeric codes applied to show how each component function supports every system level function associated with that component. A detailed listing of the codes is contained in approved plant procedures.
An understanding of the relationship between the classification system functions and component safety classification safety rules are important. This is because a system may contain components that are within the scope of the License Renewal Rule even though the host system does not perform a system level design function that would appear to be subject to the rule. It may also be because components within the system have been Page 2-3
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information credited within the CLB to do more functions than was the host system's original design functions.
Example: The Safety Injection (SI) system has as one of its intended functions, Function B, Introduce Emergency Negative Reactivity, to make the reactor subcritical. This is an obvious design function of the system as described in the UFSAR. The SI system also has selected components that are used to achieve post fire safe shutdown. That the system contains such components is not intuitive from simply reviewing the design function of the system.
Thus it is necessary to thoroughly understand and document all the uses of a system, and components within that system, in order to achieve a competent system level scoping list.
In order to differentiate between a system's design function and any other functions associated with components tracked within a system boundary, the individual License Renewal system scoping results are annotated with comments. When a system function comment describes a "primary design system function," this indicates that this is one of the designated nuclear safety functions of that system. When the comments describe "associated design system functions," this indicates a component within the system performs that function, possibly independently of what the system was originally designed to do. Primary system functions are generally discussed in the UFSAR in the chapter devoted to the system. Associated functions are generally discussed either as topical areas in the UFSAR or elsewhere in the CLB. Associated functions may or may not be nuclear safety functions.
System scoping must identify all License Renewal functions associated with components contained within a system. Generally, within the License Renewal System boundary, if the system under review contains any components that meet the License Renewal scoping criteria detailed in 10 CFR 54.4(a), the entire system is considered in-scope and that system moves forward to the License Renewal screening process.
There are two specific exceptions to this dictate:
- 1) When the only in-scope portion of the system is comprised of components that will receive a commodity group evaluation (e.g. fire barriers, equipment supports, etc.). In this case it is appropriate to identify the system or structure as not being within the scope of License Renewal, however the basis for that determination must be clearly identified.
Example:The Non-Essential Ventilation Systems contain components that act as fire barriers (fire dampers). Within the system evaluation boundary, no other functions performed by the system are License Renewal intended functions.
Therefore, this method of evaluation of the system components that perform Page 2-4
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information the fire barrier function within the Fire Barrier commodity group results in designation of the Non-Essential Ventilation Systems as not being within the scope of License Renewal.
- 2) When the only in-scope portion of the system is comprised of components that act as containment isolation boundaries. In that case it is appropriate to identify the system as not being within the scope of License Renewal so long as the components that perform the isolation boundary function are evaluated within the Containment Isolation Boundary System.
Example:The Plant Sampling System contains components that act as containment isolation boundaries (valves, pipe). Within the system evaluation boundary no components, other than those that perform the isolation function, perform any additional License Renewal intended functions. Therefore, this method of evaluation of the system components that perform the containment isolation boundary function within the Containment Isolation System results in the designation of Plant Sampling as not being within the scope of License Renewal.
The critical element of system scoping, no matter if the general results case is applied or one of the specific exceptions is invoked, is to ensure that all SSCs that perform License Renewal intended functions are identified and the criteria that made them in-scope to License Renewal is documented.
2.1.4 Design Codes, Standards, and SSC Safety Classifications The plant was designed and constructed prior to issuance of what are now the standards for Quality Group Classifications and ASME code boundaries. UFSAR Section 3.2, Classification of Structures, Components and Systems, provides a comparison of original construction codes and Regulatory Guide 1.26 and 10 CFR 50.55a. UFSAR Table 3.2-1, Classification of Structures, Systems and Components, summarizes the results of the Systematic Evaluation Program (SEP) review of selected SSCs important to safety.
The Safety Classification process provides a further comprehensive review of plant SSCs and issues Quality Classifications utilizing ANSI/ANS-51.1-1983 as guidance.
Consequently, plant SSCs have been designated as Safety Class 1 (SC-1), Safety Class 2 (SC-2), Safety Class 3 (SC-3), Safety Significant (SS) typically described at other facilities as Non-Nuclear Safety related with Augmented Quality, and Non-Nuclear Safety (NNS).
Components designated as SC-1, SC-2, or SC-3 are classified as safety related because they meet the safety related structures, system and components definition in 10 CFR 50.2.
Consequently, these components are also subject to full quality assurance requirements.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Quality group flags commensurate with safety classifications are depicted on the plant P&lDs. Components designated as SC-1, 2 or 3 are necessarily within the scope of License Renewal while other system components may or may not be.
2.1.5 Application of License Renewal Scoping Criterion Figure 2.1-1 provides a basic diagram depicting how the scoping and screening process is executed.
2.1.5.1 Safety-Related Criteria Pursuant to 10 CFR 54.4(a)(1) (Criterion 1) 10 CFR 54.4(a)(1) states that SSCs within the scope of License Renewal include safety related SSCs that are relied upon to remain functional during and following design basis events (as defined in 10 CFR 50.49(b)(1)) to ensure the following functions:
"* the integrity of the reactor coolant pressure boundary;
"* the capability to shut down the reactor and maintain it in a safe shutdown condition; or
"* the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposure comparable to those referred to in 10 CFR 50.34(a)(1), 10 CFR 50.67(b)(2), or 10 CFR 100.11 of this chapter as applicable.
As discussed in Section 2.1.4, the plant has an established safety classification process for SSCs. The safety related criterion used in that process encompasses the definition of safety related specified in 10 CFR 54.4(a)(1).
Implementation of the License Renewal Scoping and Screening procedure ensured that the UFSAR, Technical Specifications, design documents, design drawings and SSC safety classifications were reviewed as applicable to ensure all system functions were identified.
Based on this review, the License Renewal intended functions relative to the criteria of 10 CFR 54.4(a)(1) were identified and documented (as discussed in Section 2.1.3).
Thus, the scoping process used to identify safety related systems and structures is consistent with and satisfies the criteria of 10 CFR 54.4(a)(1).
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 2.1.5.2 Non-Safety Related Criteria Pursuant to 10 CFR 54.4(a)(2) (Criterion 2) 10 CFR 54.4(a)(2) states that SSCs within the scope of License Renewal include non-safety related SSCs whose failure could prevent satisfactory accomplishment of any of the functions identified for safety related SSCs.
The analytical process used to review SSCs for 10 CFR 54.4(a)(2) applicability ensured that the UFSAR, Technical Specifications, design documents, design drawings and the SSC safety classifications were reviewed as applicable to ensure all non-safety SSC functional interactions were identified where a non-safety SSC could fail and prevent the satisfactory accomplishment of a safety function.
However, reviewing design documents a system at a time does not provide information relative to system spatial interactions. Thus, it is possible that a functionally non-safety component whose failure could prevent a safety function could escape being designated as in-scope solely using the analytical review method.
To guard against such instances, the identification of SSCs meeting the criterion of 10 CFR 54.4(a)(2) was performed in a two-phase, synergistic fashion.
"* Phase one, the analytical process, evaluates SSCs meeting Criterion 2 that are explicitly identified in the Current Licensing Basis (CLB).
"* Phase two, the plant spaces physical review, evaluates SSCs for possible interactions not functionally described in the CLB.
2.1.5.3 Phase 1: Incorporating the Non-Safety SSCs Meeting the Criterion that are Explicitly Identified in the Current Licensing Basis Scoping and Screening procedures direct the review of the CLB to identify the non-safety SSCs subject to inclusion within the rule. As previously discussed, some non-safety related equipment whose failure could affect a safety related function carry a classification designator of safety significant and are included within the plant Quality Assurance Program. This quality classification grouping may also include components that fulfill 10 CFR 54.4(a)(3) criteria. As discussed in Section 2.1.3, it is possible to segregate classifications by quality criteria rule application. The quality criteria rules in approved plant procedures were reviewed to determine which SSCs met 10 CFR 54.4(a)(2). These rules allow identification of systems which host components that are:
"* Credited for HELB (pipe whip, jet impingement)
"* Credited for internal flooding (barriers, drains)
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information
"* Credited for external flooding
"* Credited for internal missiles
"* Load handling equipment credited for NUREG-0612
"* Alternate/backup systems or equipment credited in mitigating Licensing Basis Events Three items warrant further clarification and discussion: SR/NNS piping interfaces, seismic Il/I supports, and alternate/backup systems and equipment credited in mitigating Design Basis Events.
SR/NNS Piping Interface The scoping and screening process for mechanical systems utilizes P&IDs to graphically represent License Renewal boundaries. The P&IDs show component classification boundaries at valves. Actual safety boundaries extend to the weld after the first seismic support beyond the P&ID depicted class change. This piping is subject to aging management review (AMR) and is included within the characterization of the piping assets considered to be within the scope of License Renewal.
Seismic II/I Supports Non-seismic SSCs, which are positioned above or in close proximity to safety related SSCs and whose failure during a seismic event could cause the failure of a safety related SSC, are commonly referred to as Seismic Il/I.
This includes:
"* Supports for non-safety components whose hangers, supports and mounting hardware must be seismically designed to prevent non-safety related components from damaging adjacent safety related components.
"* Non-safety related components and associated supports that can be considered a potential source of jeopardy (e.g. missiles) for nearby safety related components.
NOTE: The pressure boundary aspects of non-safety piping whose failure could affect a safety function are discussed in Section 2.1.5.4.
As discussed in UFSAR Section 3.2, Classification of Structures, Components and Systems, and Section 3.7, Seismic Design, significant analysis and modification efforts have been made to ensure the station is seismically capable.
Seismic robustness was further enhanced by implementation of the Seismic Qualification Utility Group (SQUG) processes as noted in UFSAR Section Page 2-8
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 3.1.2.1.2, General Design Criterion 2, Design Basis for Protection Against Natural Phenomena. These facts notwithstanding, there is no plant labeling scheme useful for differentiating between non-safety supports that are within the scope of License Renewal and those that are not. Consequently, non-safety equipment supports in areas containing safety related equipment will be considered within the scope of License Renewal as a commodity group in accordance with procedure, and will receive subsequent aging management reviews.
Alternate/Backup Systems Credited in Mitigating Licensing Basis Events The facility was constructed prior to the issuance of the 10 CFR 50 General Design Criteria and the Standard Review Plan. As a result of the NRCs Systematic Evaluation Program (SEP) as well as closure of Unresolved Safety Issues (USIs) and Generic Safety Issues (GSIs), numerous enhancements were made to bring the facility into conformance with the intent of the more modern design criteria. One of the fundamental concepts used by the Nuclear Regulatory Commission (NRC) in assessing the acceptability of the plant design included a review of the ability to achieve safe shutdown given a postulated SSC failure and an initiating event or transient. While those events are not included in the plant's accident analysis set, they are critical elements relied upon by the NRC to judge the safety adequacy of the facility and are essential elements of the CLB. The UFSAR and NUREG-0821 describe the results of the SEP review and any modifications that were required to achieve compliance with the NRCs Safety Evaluation Reports (SERs).
Intrinsic to the CLB are the modifications or strategies used to cope with the loss of safety related SSCs from a variety of events including: Fires, Internal and External Floods, Tornadoes, High Energy Line Breaks, Internally Generated Missiles, Masonry Block Wall Failures, and Seismic Events. Modifications or coping strategies were necessary because, in some areas, the facility was not constructed with sufficient physical separation to prevent the consequences of a single initiating event from damaging both trains of equipment in a mitigating system used to satisfy a safety function. The modifications are well documented and are included within the characterization of the structural assets evaluated in the Civil or Mechanical scoping reviews as appropriate.
Where modifications alone were not an appropriate solution, other equipment sets were selected, and that equipment proven acceptable to achieve safe shutdown given the consequences of the postulated failure. The basic methodology used included ensuring that the postulated failure could not introduce any safety Page 2-9
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information consequences that could not be mitigated by surviving equipment, and identifying which mitigating equipment may be used as a surrogate or backup to the systems explicitly credited in mitigating Design Basis Events.
Example:The screen house was modified to add a dike bisecting the operating floor of the structure. The dike serves to protect vital equipment on the east side from the effects of a flood caused by a failure in the non-safety circulating water pumps on the west side. The dike is within the scope of License Renewal. The screen house can not be protected against the effects of a tornado. The plant had to modify other systems (fire water) to provide a backup cooling source for use of the emergency diesel generators to achieve safe shutdown in the event the screen house is lost. The backup cooling water is within the scope of License Renewal.
Example:Some block wall panels between the intermediate building and the turbine building are susceptible to failure in the event of high-energy line breaks or seismic events. This failure could render the preferred auxiliary feedwater system inoperable. An additional set of standby auxiliary feedwater pumps, not subject to the same failure mechanism, was installed for use in this scenario. In this example, the susceptible bock wall panels are not within the scope of License Renewal as Criterion 2 components even though the safety related preferred auxiliary feedwater pumps would be impaired by falling blocks, because the standby auxiliary feedwater pumps perform the required safety function. However, these very same wall panels are in scope as fire barriers (Criterion 3) and require aging management review appropriate to that function. In other areas of the structure, restraining/deflection devices were installed to prevent blocks from falling on safety equipment. The restraining/deflection devices are within the scope of License Renewal. The resolution of USI A-46 required the use of documented analysis and walkdowns to ensure that the systems needed to shut down the plant and maintain it in a safe condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> can withstand a design-basis seismic event. The SSCs selected for achieving safe shutdown following a design basis seismic event are within the scope of License Renewal regardless of their safety classification.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 2.1.5.4 Phase 2: Evaluation of Non-Safety SSCs for Failure Modes and Effects or Spatial Interactions not Explicitly Functionally Described in the CLB After completion of the analytical scoping effort described in Section 2.1.5.3, further evaluations are performed. In plant areas containing safety related equipment, field verifications are performed to ascertain if any systems or system piping segments not already included within the scope of License Renewal from the analytical review are present. If a newly identified system or piping segment has a failure mode or effect that meets Criterion 2, that system or segment is included within the scope of License Renewal.
Example:The evaluation of Diesel Room A shows that the room contains: fuel oil, service water, jacket water, fire water, treated water, and house heating steam. House heating steam is a high energy system and the piping and equipment segments in the diesel room could fail such that the diesel is disabled. The other fluid services in the room are included within the scope of License Renewal based on the analytical review of their functions in supporting the Diesel Generator. House heating steam must now also be included based on the Spaces Review.
Based on these reviews, the License Renewal intended functions relative to the criteria of 10 CFR 54.4(a)(2) were identified and documented (as discussed in Section 2.1.3).
Thus, the scoping process used to identify non-safety SSCs whose failure could affect a safety function is consistent with and satisfies the criteria of 10 CFR 54.4(a)(2).
2.1.5.4.1 Use of Mitigative Features or Preventive Aging Management Techniques for Non-Safety Equipment Whose Failure Could Affect a Safety Function During the review of SEP Topic 111-5.B, "Pipe Breaks Outside Containment," a comprehensive review of postulated high and moderate energy pipe breaks and cracks in accordance with BTP AJB 3-1 was performed. Postulated breaks and cracks were assumed to occur in the following systems, or portions of systems, not in the scope of license renewal per 10 CFR 54.4(a)(1) or (a)(3).
"* Circulating Water in the Screen House and Turbine Building
"* Heating Steam in the Screen House, Control Building battery rooms and relay room, Auxiliary Building and Diesel Building
"* Feedwater, Condensate, and Main Steam in the Turbine Building
"* Chilled Water in the Intermediate Building Page 2-11
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Where protective measures were needed to assure safe shutdown capability, mitigative measures were used to cope with the postulated pipe failure in all except one situation. That case is the heating steam lines in the Diesel Building.
In the former cases, the mitigative features have been placed within the scope of license renewal. In the latter case, as described in Section 2.1.5.4, all the steam heating piping and components in the Diesel Building have been placed in the scope of license renewal in accordance with the criteria specified in 10 CFR 54.4(a)(2).
Where mitigative features have been employed, those features are within the scope of license renewal and are characterized within the building that houses them. All other fluid system components designated as being within the scope of license renewal for 10 CFR 54.4(a)(2) employ preventive measures to manage aging. These system components are characterized in the system that contains them.
2.1.5.5 Other Scoping Pursuant to 10 CFR 54.4(a)(3) (Criterion 3) 10 CFR 54.4(a)(3) states that SSCs within the scope of License Renewal include all systems and structures relied on in safety analyses or plant evaluations to demonstrate compliance with the Commission's regulations for fire protection (10 CFR 50.48), environmental qualification (10 CFR 50.49), pressurized thermal shock (10 CFR 50.61), anticipated transients without scram (10 CFR 50.62), and station blackout (10 CFR 50.63).
The analytical process used to review SSCs for 10 CFR 54.4(a)(3) applicability ensured that the UFSAR, Technical Specifications, design documents, design drawings and plant safety classifications were reviewed as applicable to ensure all SSCs credited for compliance with the regulated event set are identified. Specific scoping information based on each of these regulations is described in the following sections.
2.1.5.6 Fire Protection (FP)
UFSAR Section 9.5.1, Fire Protection Systems, describes the station fire protection and post fire safe shutdown equipment. All fire protection, detection, mitigation, confinement and safe shutdown equipment used at the station is subject to Criterion 3 scoping review.
Evaluations have been made of equipment needed to meet the fire protection requirements of Appendix A to Branch Technical Position ASB 9.5-1, as well as those needed to meet 10 CFR 50, Appendix R and 10 CFR 50.48. These Page 2-12
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information evaluations are used as fire protection scoping basis documents. All structures and systems that contain components used for fire protection of the SSCs important to safety are within the scope of License Renewal.
NOTE: For commercial insurance purposes, many of the site structures not important to safety also have fire detection and mitigation. Where a fire protection system is not credited in the CLB as important to safety, that system and the SSCs it protects are not within the License Renewal Scope.
2.1.5.7 Environmental Qualification (EQ)
UFSAR Section 3.11, Environmental Design of Mechanical and Electrical Equipment, describes the station's process for Environmental Qualification (EQ).
The master list of EQ components is detailed in site-specific procedures. All systems that contain components detailed on the EQ master equipment list are within the scope of License Renewal.
2.1.5.8 Pressurized Thermal Shock (PTS)
UFSAR Section 5.3.3.5, Pressurized Thermal Shock, describes the stations conformance to the 10 CFR 50.61 criterion.
RG&E has made two submittals to the NRC regarding Pressurized Thermal Shock (PTS). The first was made on January 23, 1986 in response to the original version of 10 CFR 50.61. The NRC SE on this submittal was issued on March 11, 1987 and concluded that the calculated reference temperature for pressurized thermal shock (RTPTS) met 10 CFR 50.61 requirements. When 10 CFR 50.61 was amended on May 15, 1991, RG&E made a second submittal on February 13, 1992. The NRC dispositioned this submittal in their SER on the amendment to recapture the construction period for Ginna dated April 20, 1994. In this SER, the NRC found that the reference temperature found for pressurized thermal shock for the Ginna reactor vessel will be well below the screening criteria at the expiration of its current license. The RG&E submittals and NRC SERs did not identify the need for specific plant hardware modifications or reliance on other plant systems.
Consequently only the reactor vessel is credited for PTS. Therefore, the only system relied upon for PTS is the Reactor Coolant System.
2.1.5.9 Anticipated Transients Without a Scram (ATWS)
UFSAR Section 7.2.6, Anticipated Transients Without a Scram Mitigation System Actuation Circuitry, describes the system installed to provide conformance with 10 CFR 50.62. All equipment in the system from the sensor output to the final Page 2-13
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information actuation device that is credited for compliance with 10 CFR 50.62 is included in the scope of License Renewal. All systems that host those components are within scope of License Renewal.
2.1.5.10 Station Blackout (SBO)
UFSAR Section 8.1.4.5, Station Blackout Program, describes the station's methodology for coping with a station blackout. The coping strategy basis reference documents include both primary and alternative SSCs available to manage the event. Certain plant areas may heat up from SBO-induced loss of ventilation. Where heatup calculations take credit for structural elements as a heat sink, that structure is also considered a primary SBO mitigating component. The primary mitigating SSCs are within the scope of License Renewal. License Renewal regulatory guidance also mandates the inclusion of the plant system portion of the offsite power SSCs used for SBO recovery beyond those identified in the regulatory commitments made to satisfy 10 CFR 50.63 criteria. Systems and structures that provide a function for SBO coping and systems or structures that provide a function for recovery from an SBO condition in accordance with the current License Renewal regulatory interpretation are within the scope of License Renewal.
Based on these reviews, the License Renewal intended functions relative to the criteria of 10 CFR 54.4(a)(3) were identified and documented (as discussed in Section 2.1.3).
Thus, the scoping process used to identify systems and structures relied upon to mitigate the regulated events of concern is consistent with and satisfies the criteria of 10 CFR 54.4(a)(3).
2.1.6 Interim Staff Guidance Discussion Recently during license renewal application reviews the NRC staff identified six issues for which additional staff and industry guidance clarification was necessary. They are:
- 1. Housing of Active Components
- 2. Concrete Aging Requirements
- 3. Interpretation of 10 CFR 54(a)2
- 4. Fire Protection Component Aging Management
- 5. Treatment of Electrical Fuse Holders
- 6. Scoping of Station Blackout Components Page 2-14
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Following is a discussion of the general process used during the License Renewal Integrated Plant Assessment at Ginna Station for each of these areas:
Housing of Active Components The Statements of Consideration for 10 CFR 54 provides the License Renewal Rule philosophy that, during the extended period of operation, safety-related functions should be maintained in the same manner and to the same extent as during the current licensing term.
Examples of structures and components that perform passive functions are listed in 10 CFR 54.21 (a)(1)(ii), which states, "These structures and components include, but are not limited to, pump casings, valve bodies..."
Pumps and valves were just an example here, meant to focus the AMR process on the passive function of an SSC. That passive function is not limited to the pressure boundary of the reactor coolant system. The exclusion of an SSC due to its active nature only applies to that portion of the SSC with an active function and not to those portions of the SSC with a passive function. Therefore, at Ginna Station, fan housings and fire damper housings are considered to be within scope and subject to an AMR.
Concrete Aging Requirements As a result of the performance of AMRs for in-scope concrete components, Ginna Station has concluded that many of these components do not require aging management for the period of extended operation. This conclusion is based on a review of the material of construction, the environment, and industry and plant-specific operating experience for these components. However, for accessible concrete portions of the containment, Ginna Station has implemented the examination requirements and inspection intervals of ASME Section Xl, Subsection IWL as an aging management program (AMP) for the period of extended operation. Other structures subject to an aging management review will receive similar inspections as part of the Structures Monitoring Program. The GALL report does not recommend further evaluation of concrete components in inaccessible areas for which the applicant can demonstrate a non-aggressive environment. The environment evident for the inaccessible concrete at Ginna Station is not aggressive; therefore, further evaluation of normally inaccessible structures will only be considered when excavations allow access or when aging effects on accessible concrete structures indicate that potential detrimental aging effects could also be occurring in inaccessible areas.
Interpretation of 10 CFR 54(a)2 10 CFR 54.4(a)(2) states that SSCs within the scope of License Renewal include non-safety related SSCs whose failure could prevent satisfactory accomplishment of any of the functions identified for safety related SSCs.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information The analytical process that was used to review the Ginna Station SSCs for 10 CFR 54.4(a)(2) applicability ensured that the UFSAR, Technical Specifications, design documents, design drawings and the SSC safety classifications were reviewed as applicable to identify all non-safety SSC functional interactions where a non-safety SSC could fail and prevent the satisfactory accomplishment of a safety function.
However, reviewing design documents a system at a time did not provide information relative to system spatial interactions. Thus, it was possible that a functionally non-safety component whose failure could prevent a safety function could escape being designated as in-scope solely using the analytical review method.
To guard against such instances, the identification of SSCs meeting the criterion of 10 CFR 54.4(a)(2) was performed using a two-phase approach. Phase one was an analytical process, which evaluated SSCs meeting Criterion 2 that were explicitly identified in the Current Licensing Basis (CLB). Phase two was a plant spaces physical review, which evaluated SSCs for possible interactions that were not explicitly described in the CLB.
The result was synergistic in that the 54.4(a)(2) criterion was applied such that non-safety related SSCs were identified as being within the scope of license renewal when there was a potential either physically or spatially for interaction with the intended function of safety related equipment. A more detailed discussion of this phased methodology is available in Section 2.1.5.3 and Section 2.1.5.4.
Fire Protection Component Aging Management In a January 28, 2002, letter from the NRC to NEI entitled, "Proposed Staff Guidance On Aging Management Of Fire Protection Systems For License Renewal," the NRC provided changes to their previous guidance to the industry on aging management for passive SSCs comprising fire protection. As outlined in the program description in Section 2.2.10, Fire Protection Program, and Section 2.2.11, Fire Service Water System Program, Ginna Station intends to provide for aging management in a manner consistent with the proposed guidance.
This includes such guidance as to the performance of volumetric inspections and/or wall thickness evaluations, and visual as well as other techniques. For the Fire Water System Program, Ginna Station will continue to flow test those portions of the sprinkler system that are now routinely tested.
Treatment of Electrical Fuse Holders Consistent with the requirements specified in 10 CFR 54.4(a), fuse holders (including fuse clips and fuse blocks) are considered to be passive electrical components. Fuse holders are scoped, screened, and included in the aging management review (AMR) in the same manner as terminal blocks and other types of electrical connections. However, fuse holders Page 2-16
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information inside the enclosure of an active component, such as switchgear, power supplies, power inverters, battery chargers, and circuit boards, are considered to be piece parts of the larger assembly. Since piece parts and subcomponents in such an enclosure are inspected regularly and maintained as part of the Ginna Station normal maintenance and surveillance activities, they are considered not subject to an AMR. Fuse holders perform a primary function similar to electrical connections by providing an electrical circuit to deliver rated voltage, current, or signals. These intended functions meet the criteria of 10 CFR 54.4(a).
Additionally, these intended functions are performed without moving parts or without a change in configuration or properties as described in 10 CFR 54.21 (a)(1)(i). Fuse holders are therefore passive, long-lived electrical components within the scope of license renewal and subject to an AMR. Therefore, aging management of the fuse holders is required for those cases where fuse holders are not considered piece parts of a larger assembly.
Scoping of Station Blackout Components NRC guidance on this issue is as follows: "Consistent with the requirements specified in 10 CFR 54.4(a)(3) and 10 CFR 50.63 (a)(1), the plant system portion of the offsite power system should be included within the scope of license renewal." Further clarification was provided which stated that, "the staff has determined that the plant system portion of the offsite power system that is used to connect the plant to the offsite power source should be included within the scope of the rule. This path typically includes the switchyard circuit breakers that connect to the offsite power system transformers (Startup transformers), the transformers themselves, the intervening overhead or underground circuits between circuit breaker and transformer and transformer and onsite electrical distribution system, and the associated control circuits and structures."
UFSAR Section 8.1.4.5, Station Blackout Program, describes the station's methodology for coping with a station blackout. The SSCs assumed to be necessary for the coping strategy including both primary and alternative SSCs available to manage the event are included within the scope of License Renewal.
As stated above the License Renewal regulatory guidance also mandates the inclusion of selected offsite power SSCs used for SBO recovery beyond those identified in the regulatory commitments made to satisfy 10 CFR 50.63 criteria. Therefore, systems and structures that provide a function for SBO coping and systems or structures that provide a function for recovery from an SBO condition in accordance with the current License Renewal regulatory interpretation are also considered within the scope of License Renewal.
Additional specific information on this methodology is included in Section 2.1.5.10, Station Blackout.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 2.1.7 Component Level Screening (Identification of Components Subject to Aging Management Review)
The requirement to identify SCs subject to an aging management review is specified in 10 CFR 54.21 (a)(1) that states:
"Each application must contain the following information:
(a) An integrated plant assessment (IPA). The IPA must:
(1) For those systems, structures, and components within the scope of this part, as delineated in 10 CFR 54.4, identify and list those structures and components subject to an aging management review. Structures and components subject to an aging management review shall encompass those structures and components:
(i) That perform an intended function, as described in 10 CFR 54.4 without moving parts or without a change in configuration or properties. These structures and components include, but are not limited to, the reactor vessel, the reactor coolant system pressure boundary, steam generators, the pressurizer, piping, pump casings, valve bodies, the core shroud, component supports, pressure retaining boundaries, heat exchangers, ventilation ducts, the containment, the containment liner, electrical and mechanical penetrations, equipment hatches, seismic Category 1 structures, electrical cables and connections, cable trays, and electrical cabinets, excluding, but not limited to, pumps (except casing), valves (except body), motors, diesel generators, air compressors, snubbers, the control rod drive, ventilation dampers, pressure transmitters, pressure indicators, water level indicators, switchgears, cooling fans, transistors, batteries, breakers, relays, switches, power inverters, circuit boards, battery chargers, and power supplies; and (ii) That are not subject to replacement based on a qualified life or specified time period."
The screening portion of the IPA methodology is divided into three engineering disciplines; mechanical, civil/structural, and electrical/l&C. The relevant aspects of the component/structural component scoping and screening process for mechanical systems, civil structures, and electrical and I&C systems are described below.
For mechanical systems and civil structures, this process establishes evaluation boundaries, determines the SCs that compose the system or structure, determines which of those SCs support system/structure intended functions, and identifies specific SC intended functions. Consequently, not all of the SCs for in-scope systems or structures are in the Page 2-18
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information scope of license renewal. Once the in-scope SCs are identified, the process determines which SCs are subject to an aging management review per the criteria of 10 CFR 54.21 (a)(1). Note that the screening process is consistent with the NRC Staff's guidance on consumables provided in Table 2.1-3 of NUREG-1800 (Reference 6).
For electric/l&C systems, a bounding "Plant Spaces" approach as described in NEI 95-10 is taken. This approach establishes evaluation boundaries, determines the electrical and I&C component commodity groups within in-scope systems, identifies specific component and commodity-intended functions and then determines which component commodity groups are subject to an aging management review per the criteria of 10 CFR 54.21 (a)(1). This approach calls for component scoping after screening has been performed.
2.1.7.1 Mechanical Systems For mechanical systems, the component/structural component scoping and screening process is performed on each system identified to be within the scope of license renewal. This process evaluates the individual SCs included within in-scope mechanical systems to identify specific SCs or SC groups that require an aging management review. Electrical interface components associated with mechanical systems that are determined to be in scope are evaluated as described in Section 2.1.7.4, Electrical and I&C Systems.
Mechanical system evaluation boundaries were established for each system within the scope of license renewal. These boundaries were determined by mapping the pressure boundary associated with license renewal system intended functions onto the system flow diagrams. License renewal system intended functions are the functions a system must perform relative to the scoping criteria of 10 CFR 54.4(a)(1), 10 CFR 54.4(a)(2), and 10 CFR 54.4(a)(3).
The sequence of steps performed on each mechanical system determined to be within the scope of license renewal is as follows:
"* Based on a review of design drawings and the system component list from the CMIS database, SCs that are included within the system are identified.
"* Based on the plant level scoping results, the pressure boundary associated with license renewal system intended functions is mapped onto the system's flow diagrams.
NOTE: The LR evaluation markups for a system have typically been extended to the first normally closed manual valve, check valve or automatic valve that gets a signal to go closed. A normally open manual valve has also been used as a boundary in a few instances where a failure downstream of the valve has no short Page 2-19
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information term effects, can be quickly detected, and the valve can be easily closed by operators to establish the pressure boundary prior to any adverse consequences.
However, for SBO, Appendix R, high energy line break (HELB), and flooding events, the LR boundaries for a system have been defined consistent with the boundaries established in the CLB evaluations. Those boundaries do not always coincide with an isolation device. As described in Section 2.1.5.3, actual aging management evaluation boundaries may extend beyond the graphical depiction of the screening boundary.
"* The system SCs that are within the scope of license renewal (i.e., required to perform a license renewal system intended function) are identified.
"* Component intended functions for in-scope SCs are identified. The component intended functions identified are based on the guidance of NEI 95-10.
"* The in-scope SCs that perform an intended function without moving parts or without a change in configuration or properties (screening criterion of 10 CFR 54.21 (a)(1)(i)) are identified. Active/passive screening determinations are based on the guidance in Appendix B to NEI 95-10.
" The passive, in-scope SCs that are not subject to replacement based on a qualified life or specified time period (screening criterion of 10 CFR 54.21 (a)(1)(ii)) are identified as requiring an aging management review. The determination of whether passive, in-scope SC has a qualified life or specified replacement time period was based on a review of plant-specific information, including the component database, maintenance programs, and procedures.
Example:Some of the ventilation systems in the scope of license renewal include system filters such as fiberglass prefilter elements, HEPA filters and charcoal filters. These system filters are also in the scope of license renewal, but are periodically replaced and are not subject to aging management review. Periodic testing and inspection programs are in place to monitor filter performance such that system intended functions are maintained.
2.1.7.2 Civil Structures For structures, the component/structural component scoping and screening process is performed on each structure identified to be within the scope of license renewal. This method evaluates the individual SCs included within in-scope structures to identify specific SCs or SC groups that require an aging management review.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information The sequence of steps performed on each structure determined to be within the scope of license renewal is as follows:
"* Based on a review of design drawings, the structure component list from the CMIS database and plant walkdowns, SCs that are included within the structure are identified. These SCs include items such as walls, pipe and equipment supports, conduit, cable trays, electrical enclosures, instrument panels, and related supports.
"* The CLB is reviewed and compared to the walkdown results. Appurtenances such as flood barriers, missile shields, jet impingement shields, etc., relied upon in the licensing basis are verified as accounted for within a structure.
"* The SCs that are within the scope of license renewal (i.e., required to perform license renewal system intended functions) are identified.
"* Component intended functions for in-scope SCs are identified. The component intended functions identified are based on the guidance of NEI 95-10.
"* The in-scope SCs that perform an intended function without moving parts or without a change in configuration or properties (screening criterion of 10 CFR 54.21 (a)(1)(1)) are identified. Active/passive screening determinations are based on the guidance in Appendix B to NEI 95-10.
"* The passive, in-scope SCs that are not subject to replacement based on a qualified life or specified time period (screening criterion of 10 CFR 54.21 (a)(1)(ii)) are identified as requiring an aging management review. The determination of whether a passive, in-scope SC has a qualified life or specified replacement time period was based on a review of plant-specific information, including the component database, maintenance programs and procedures, vendor manuals, and plant experience.
NOTE: The structural component function(s) may support the intended function(s) of the structure or may have a unique function that does not support the intended function of the structure. A case in point is the spent fuel storage racks that are located in the Auxiliary Building. A unique function of the spent fuel storage racks is to maintain separation of the fuel assemblies to prevent criticality, which is not considered to be an intended function of the Auxiliary Building itself. The spent fuel racks are evaluated in the Spent Fuel Cooling and Fuel Storage System.
Structural steel, anchor bolts, base plates, etc. that are required to support non safety related components to prevent physical interactions with safety related equipment are subject to aging management reviews. These components must Page 2-21
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information remain in place such that they do not impact equipment that is required to perform a safety function in such a way as to prevent the equipment from performing its safety function.
Materials such as caulking and waterstops are identified generically. Limited situations may exist where these materials are important to maintaining the integrity of the components to which they are connected. The license renewal structure or component intended functions supported by these materials are restricted to two functions. These functions are:
- 1) Providing a rated fire barrier.
- 2) Providing a flood barrier.
Sealants and caulking that support the fire barrier function are addressed as part of the fire barrier penetration seals. Waterstops that support the flood barrier function are addressed with the wall or floor within which the sealant/waterstop is contained. Flood barriers are addressed in the buildings that contain them.
2.1.7.3 Structural Commodity Groups Civil structures within the scope of License Renewal also house and support functionally unique features that may be included within the scope of the rule.
These structural elements are best described and evaluated within structural commodity groupings.
Example:The Auxiliary Building contains hundreds of fire barriers and seals. Most of these barriers perform a license renewal intended function. Rather than listing all of the barrier numbers as a subset of the Auxiliary Building components, the barriers are binned together in a commodity group and structural screening is performed on the group.
The structural commodity screening evaluation groups are:
"* Fire doors, barrier penetration seals and wraps
"* Racks, panels and electrical enclosures, pipe and equipment supports.
2.1.7.4 Electrical and I&C Systems Screening for electrical/!I&C components is performed on a generic component commodity group basis for all electrical/l&C systems, as well as the electrical/l&C component commodity groups associated with in-scope mechanical systems and civil structures. The methodology employed is consistent with the guidance in NEI 95-10 and NUREG-1 800. Components within the scope of 10 CFR 50.49 (Environmental Qualification) are not subject to an aging management review Page 2-22
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information (AMR) based on the screening criteria of 10 CFR 54.21(a)(1)(ii). This is supported by NUREG-1 800, Section 2.5.3. Additionally, each electrical and I&C system received a system level function review as described in Section 2.1.3. Although not necessary for determining the population of the passive long lived electrical components subject to aging management review, the system level review provides useful documentation of a systems role within the current licensing basis.
Screening for electrical and I&C system commodity groups used the plant spaces approach and the bounding review technique. Using this methodology, initially all passive long lived electrical and I&C commodity groups are considered subject to an AMR. The plant is segregated into areas where common, bounding environmental parameters can be assigned. Electrical commodity groups that have been installed in the area are identified and, among those, the commodity that represents the limiting aging characteristics is identified. The selected commodity is compared to the plant space service conditions and an assessment is made as to whether this limiting material will be able to maintain its function for the period of extended operation (i.e. receives an AMR). The results and conclusions of the AMR may indicate that aging management activities are required for components with specific material/environment combinations. Using these results, component specific scoping may be performed to limit the number of components for which aging management activities are required, or eliminate aging management activities altogether if nothing remains in the material/environment group population. Component scoping is performed by determining if a device served by the representative material commodity group has a License Renewal intended function and determining if credible failure modes would result in a loss of intended function. The intended function of the device being served can usually be determined by reviewing the intended functions of the devices being interconnected by the commodity group.
During electrical screening, passive, long-lived electrical/l&C sub-components that are part of larger components are evaluated as part of the larger component.
For example, wires and connections internal to relays, relay racks, motors, motor control centers, and control boards are not considered as subject to aging management review. The sub-components are part of the active component and therefore are not evaluated separately. The boundary components for the electrical/l&C component review are the incoming 34.5 kV switchyard bus breakers and the generator step up transformer. These reference points represent the transition from site controlled power systems to the power systems maintained as part of the local distribution grid. All other components associated with offsite Page 2-23
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information substations are not included in the evaluation boundary. This does not immediately exclude the on-site transformer yard and associated component commodity groups.
Based on a review of previous license renewal applications, NEI 95-10, and standard industry guidance for aging evaluation of electrical commodities, the following list represents the passive electrical/l&C component commodity groups:
- Insulated Cables and Connections (including splices, connectors, terminal blocks and fuse holders)
- Electrical Penetration Assemblies
- Phase Bus
- Switchyard Bus
- Transmission Conductors
- Uninsulated Ground Conductors
- High Voltage Insulators A review of the UFSAR, the plant's database, and design basis documents was performed to validate the commodity group applicability to Ginna Station. The review did not identify a need to add any additional groups. For these commodities, the intended function is to electrically connect a specified section of an electrical circuit to deliver voltage, current or signal or, in the case of high voltage insulators, to insulate and support an electrical conductor.
Because the electrical screening and scoping is iterative in nature, the electrical and I&C commodities above are initially considered subject to aging management review.
2.1.7.5 Screening of Stored Equipment In response to the NRC letter from Chris Grimes to Doug Walters (NEI) dated February 11, 1999,
Subject:
Screening of Equipment Kept in Storage, a review has been performed to identify equipment that (1) is maintained in storage, (2) is reserved for installation in the plant in response to a design basis event (DBE),
and (3) requires an AMR. In addition to passive components, the review has also considered stored active components that are not routinely inspected, tested, and maintained.
Page 2-24
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information The Appendix R stored equipment is used to restore power to pre-selected plant components and to provide cooling to certain areas after a fire in order to attain cold shutdown. The stored equipment identified as requiring an aging management review is cable and connectors.
2.1.7.6 Screening of Thermal Insulation In response to NRC staff requests for additional information (RAI) on other license renewal applications, a screening review has been performed of thermal insulation. The review has concluded that only the thermal insulation used on the primary containment liner is included within the scope of License Renewal.
2.1.7.7 Identification of Short-lived Components and Consumables The screening process has identified the passive components and structural members within the scope of license renewal. That process, as described in Section 2.1.7, has not attempted to identify those components that can be treated as short-lived and, therefore, do not require an AMR. These determinations have been made during the AMR process. It was during this part of the process that the plant procedures being credited for managing the effects of aging have been reviewed. If a procedure was found to provide for the periodic replacement of the component, or the component was found to have an established qualified life (e.g., for EQ purposes), the component has been identified as short-lived and an aging management review has not been required for that component.
Consumables are a special class of short-lived items that can include packing, gaskets, component seals, 0-rings, oil, grease, component filters, system filters, fire extinguishers, fire hoses, and air packs. Many types of consumables are part of a component such as a valve or a pump and, therefore, have not been identified during screening. Items potentially treatable as consumables have been evaluated consistent with the information presented in Table 2.1-3 of Reference 6.
The results of that evaluation are presented below.
2.1.7.7.1 Packing, Gaskets, Component Seals, and O-Rings Packing, gaskets, component mechanical seals, and 0-rings are typically used to provide a leak-proof seal when components are mechanically joined together. These items are commonly found in components such as valves, pumps, heat exchangers, ventilation units/ducts, and piping segments. These types of consumables are considered subcomponents of the identified components and, therefore, are not subject to their own condition or performance monitoring. Therefore, the AMR for the component has included Page 2-25
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information an evaluation of the sealing materials in those instances where it could not be demonstrated that one of the following conditions exist:
- 1. The sealing materials are short-lived because they are replaced on a fixed frequency or have a qualified life established (e.g., for EQ purposes), or
- 2. The sealing materials are not relied on in the CLB to maintain any of the following:
- Leakage below established limits
- System pressure high enough to deliver specified flow rates
- A pressure envelope for a space Note: Sealants used to provide flood and fire barriers are addressed in Section 2.1.7.2.
2.1.7.7.2 Oil, Grease, and Filters Oil, grease, and filters (both system and component filters) have been treated as consumables because either:
"* A program for periodic replacement exists, or
"* A monitoring program (e.g., predictive analysis activities, condition monitoring) exists that replaces these consumables, based on established performance criteria, when their condition begins to degrade but before there is a loss of intended function.
2.1.7.7.3 Fire Extinguishers, Fire Hoses, and Air Packs Components such as fire hoses, fire extinguishers, self-contained breathing apparatus (SCBA), and SCBA cylinders are considered to be consumables and are routinely tested or inspected. The Fire Protection Program complies with the applicable safety standards (NFPA-10 for fire extinguishers; NFPA-1 962 for fire hoses; 42 CFR 84, 29 CFR 19.10, 29 CFR 19.26, NUREG-41, and ANSI-Z88.2 for air packs), which specify performance and condition monitoring programs for these specific components. Fire hoses and fire extinguishers are inspected and hydrostatically tested periodically and must be replaced if they do not pass the test or inspection. SCBA and SCBA cylinders are inspected and periodically tested and must be replaced if they do not pass the test or inspection. The Fire Protection Program determines the replacement criterion of these components that are routinely checked by tests or inspections to assure operability. Therefore, while these consumables are in the scope of license renewal, they do not require an AMR.
Page 2-26
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Section 2.1 References
- 1. Regulatory Guide 1.97, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.
- 2. NEI 95-10, Industry Guideline for Implementing the Requirements of 10 CFR Part 54
-The License Renewal Rule, Rev. 3, Nuclear Energy Institute, March 2001.
- 3. Branch Technical Position (BTP) APCSB 9.5-1, Appendix A, Guidelines for Fire Protection for Nuclear Power Plants August 23, 1976.
- 4. Letter of August 5, 1999 from Christopher I. Grimes of the NRC to Douglas J. Walters of NEI,
Subject:
License Renewal Issue No. 98-0082, Scoping Guidance.
- 5. Letter of February 11, 1999 from Christopher I. Grimes of the NRC to Doug Walters of NEI,
Subject:
Request for Additional Information Regarding Generic License Renewal Issue No. 98-0102, Screening of Equipment that is Kept in Storage.
- 6. Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, NUREG-1 800, U.S. Nuclear Regulatory Commission, July 2001.
Page 2-27
Table 2.1-1 System Function Codes System r Function Description ANSI 51.1 Notes 10 CFR 54.4(a)
Function Discussion Criterion Code Paragraph A MAINTAIN REACTOR CORE ASSEMBLY 4.1.1,4.4.1 GEOMETRY B INTRODUCE EMERGENCY NEGATIVE 4.2.1,4.8.1 Functions include limiting the introduction of 1 REACTIVITY TO MAKE THE REACTOR positive reactivity.
SUBCRITICAL C INTRODUCE NEGATIVE REACTIVITY TO 4.2.1 1 ACHIEVE OR MAINTAIN SUBCRITICAL REACTOR CONDITION D SENSE OR PROVIDE PROCESS CONDITIONS 4.3.1 1 AND GENERATE SIGNALS FOR REACTOR TRIP AND ENGINEERED SAFETY FEATURES ACTUATION E PROVIDE REACTOR COOLANT PRESSURE 4.4.1 1 BOUNDARY F REMOVE RESIDUAL HEAT FROM THE RCS 4.5.1 Residual heat removal by direct recirculation of 1 reactor coolant. This system function does not address emergency core cooling via Engineered Safety Features Actuation.
G PROVIDE EMERGENCY CORE COOLANT 4.8.1 This function includes coolant that is provided to 1 WHERE THE ECCS PROVIDES COOLANT the core via RCS piping. This function also DIRECTLY TO THE CORE addresses coolant inventory that is maintained for use by the ECCS to provide emergency core cooling and to introduce negative reactivity.
Page 2-28 Application for Renewed Operating License
Table 2.1 -1 System Function Codes System Function Description ANSI 51.1 Notes 10 CFR 54.4(a)
Function Discussion Criterion Code Paragraph H PROVIDE EMERGENCY HEAT REMOVAL 4.10.1 Secondary heat removal capability addresses the 1 FROM THE REACTOR COOLANT SYSTEM secondary side of the steam generators, and USING SECONDARY HEAT REMOVAL steam generator secondary cooling paths, for CAPABILITY example: Auxiliary Feedwater, Relief Valves/Lines.
J PROVIDE HEAT REMOVAL FROM SAFETY 4.7.1 This function addresses heat removal as 1 RELATED HEAT EXCHANGERS necessary to provide a nuclear safety function.
This function does not include emergency heat removal from the secondary side of the Steam Generators (System Function H).
K PROVIDE PRIMARY CONTAINMENT 4.9.1 This function addresses any primary containment 1 BOUNDARY fission product barrier or primary containment radioactive material holdup or isolation.
L PROVIDE EMERGENCY HEAT REMOVAL 4.11.1 1 FROM PRIMARY CONTAINMENT AND PROVIDE CONTAINMENT PRESSURE CONTROL M PROVIDE EMERGENCY REMOVAL OF 4.11.1 RADIOACTIVE MATERIAL FROM THE PRIMARY CONTAINMENT ATMOSPHERE N CONTROL COMBUSTIBLE GAS MIXTURES IN 4.11.1 THE PRIMARY CONTAINMENT ATMOSPHERE Page 2-29 Application for Renewed Operating License
Table 2.1-1 System Function Codes System Function Description ANSI 51.1 Notes 10 CFR 54.4(a)
Function Discussion Criterion Code Paragraph O MAINTAIN EMERGENCY TEMPERATURES 4.12.1 1 WITHIN AREAS CONTAINING SAFETY CLASS 1,2,3 COMPONENTS P ENSURE ADEQUATE COOLING INTHE SPENT 4.13.1 Cooling to maintain stored fuel within acceptable 1 FUEL POOL temperature limits.
o PROVIDE ELECTRICAL POWER TO SAFETY 4.14.1 1 CLASS 1,2,3 COMPONENTS R STRUCTURALLY SUPPORT OR HOUSE 4.18.1 1 SAFETY CLASS 1,2,3 COMPONENTS S SPECIAL CAPABILITY CLASS FUNCTIONS N/A Components within the system are Safety N/A Significant (augmented quality). For the purposes of License Renewal, components which are special capability class are treated under the Criterion 3 Codes Z1 through Z5.
T NON-NUCLEAR SAFETY CLASS FUNCTIONS N/A N/A X SFR FUNCTION NOT APPLICABLE AT N/A Not a system level function. Component performs 1 COMPONENT LEVEL a safety related function beyond the boundaries of the respective system specific design, such as accident monitoring.
Page 2-30 Application for Renewed Operating License
Table 2.1-1 System Function Codes System Function Description ANSI 51.1 Notes 10 CFR 54.4(a)
Function Discussion Criterion Code Paragraph Y LICENSE RENEWAL CRITERION 2 - NON N/A 2 SAFETY RELATED SSCs WHOSE FAILURE COULD PREVENT SATISFACTORY ACCOMPLISHMENT OF A SAFETY RELATED FUNCTION Zi LICENSE RENEWAL CRITERION 3 - SSCs N/A 3 RELIED UPON IN SAFETY ANALYSES OR PLANT EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR FIRE PROTECTION (10 CFR 50.48)
Z2 LICENSE RENEWAL CRITERION 3 - SSCs N/A 3 RELIED UPON IN SAFETY ANALYSES OR PLANT EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR ENVIRONMENTAL QUALIFICATION (10 CFR 50.49)
Z3 LICENSE RENEWAL CRITERION 3 - SSCs N/A 3 RELIED UPON IN SAFETY ANALYSES OR PLANT EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR PRESSURIZED THERMAL SHOCK (10 CFR 50.61)
Page 2-31 Application for Renewed Operating License
Table 2.1-1 System Function Codes System Function Description ANSI 51.1 Notes 10 CFR 54.4(a)
Function Discussion Criterion Code Paragraph Z4 LICENSE RENEWAL CRITERION 3 - SSCs N/A 3 RELIED UPON IN SAFETY ANALYSES OR PLANT EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR ANTICIPATED TRANSIENTS WITHOUT SCRAM (10 CFR 50.62)
Z5 LICENSE RENEWAL CRITERION 3 - SSCs N/A 3 RELIED UPON IN SAFETY ANALYSES OR PLANT EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR STATION BLACKOUT (10 CFR 50.63)
Page 2-32 Application for Renewed Operating License
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Figure 2.1-1 Scoping and Screening Process Overview Systems, structures and their functions identified from UFSAR, design drawings Component Quality List (Q-List),
Design Documents, and other design documents.
System and structure boundaries defined and component population lists established.
System and structure functions compared against criteria in 54.4(a)(1)-(3).
I, Plant spaces evaluated to identify systems/components whose failure can lead to a loss of safety function (54.4(a)(2) physical review). iI Systems and structures with functions meeting criteria in Systems and structures with no functions 54.4(a)(l)-(3) included within the scope of license renewal. meeting criteria in 54.4(a)(1)-(3) removed from scope.
I In-scope system and structure components and component groups identified from Q-list, drawings, design documentation and plant walkdowns.
Structures and components in scope reviewed to identify those subject to aging management review.
- Passive
- Long-lived
- Support system intended function Structures and components not required to support system intended functions and active components not subject to AMR.
List of structures and components requiring aging management review prepared with associated intended functions.
Page 2-33
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 2.2 Plant Level Scoping Results The systems, structures, and commodities at Ginna Station were evaluated as to whether they were within the scope of license renewal, using the methodology described in Section 2.1. The results are shown below in Table 2.2-1.
Page 2-34
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.2-1 Plant Level Scoping Results Description Within Scope of Comments License Renewal?
SRP Evaluation Group: Reactor Vessel, Internals, and Reactor Coolant System Reactor Coolant, Core, and Internals Yes The Regenerative and Letdown Heat (Section 2.3.1.1) Exchangers are included in this system.
Non-Class 1 RCS Components Yes (Section 2.3.1.6)
SRP Evaluation Group: Engineered Safety Features Containment Hydrogen Detectors and Yes Recombiners (Section 2.3.2.4)
Containment Isolation Components Yes Piping and components in the Heating (Section 2.3.2.5) Steam System, Plant Air Systems, Plant Sampling Systems, and Containment Structure that are associated with the primary containment boundary are included in CICS.
Containment Spray (Section 2.3.2.2) Yes Residual Heat Removal (Section Yes 2.3.2.3)
Safety Injection (Section 2.3.2.1) Yes SRP Evaluation Group: Auxiliary Systems Chemical and Volume Control (Section Yes The Regenerative and Letdown Heat 2.3.3.1) Exchangers are included in the Reactor Coolant, Core, and Internals system.
Chilled Water (Section 2.3.3.15) No Page 2-35
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.2-1 Plant Level Scoping Results Description Within Scope of Comments License Renewal?
Circulating Water (Section 2.3.3.14) No Those portions of the CW system that support the delivery of lake water sufficient for the use of service water and fire water pumps are evaluated within the Service Water System. The interface with the Circulating Water System that provides circulating water flood detection is evaluated within the Reactor Protection System.
Component Cooling Water (Section Yes 2.3.3.2)
Containment Ventilation Systems Yes (Section 2.3.3.9)
Cranes, Hoists, and Lifting Devices Yes NUREG-0612 cranes are included in this (Section 2.3.3.11) group.
Essential Ventilation Systems (Section Yes 2.3.3.10)
Emergency Power (Section 2.3.3.8) Yes Fire Protection (Section 2.3.3.6) Yes Fire protection commodity group items are included within this system.
Fuel Handling (Section 2.3.3.16) No Cranes, new and spent fuel storage racks, the spent fuel pool and cavity liners are evaluated separately.
Heating Steam (Section 2.3.3.7) Yes Piping and components associated with the primary containment boundary are included with CICS.
Non-Essential Ventilation Systems No Components that act as fire barriers (fire (Section 2.3.3.19) dampers) are included in the Fire Barrier commodity group.
Plant Air Systems (Section 2.3.3.18) No Piping and components associated with the primary containment boundary are included with CICS.
Page 2-36
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.2-1 Plant Level Scoping Results Description Within Scope of Comments License Renewal?
Plant Sampling Systems (Section No Piping and components associated with 2.3.3.17) the primary containment boundary are included with CICS.
Radiation Monitoring (Section Yes 2.3.3.13)
Service Water (Section 2.3.3.5) Yes Those portions of the CW system that support the delivery of lake water sufficient for the use of service water and fire water pumps are evaluated within the Service Water System.
Site Service and Facility Support No (Section 2.3.3.20)
Spent Fuel Cooling and Fuel Storage Yes New and spent fuel storage racks, the (Section 2.3.3.3) spent fuel pool, transfer tube, and cavity liners are included in this group.
Treated Water (Section 2.3.3.12) Yes Waste Disposal (Section 2.3.3.4) Yes SRP Evaluation Group: Steam and Power Conversion System Auxiliary Feedwater (Section 2.3.4.3) Yes Feedwater and Condensate (Section Yes 2.3.4.2)
Main and Auxiliary Steam (Section Yes 2.3.4.1)
Turbine-Generator and Supporting Yes Systems (Section 2.3.4.4)
SRP Evaluation Group: Containments, Structures and Component Supports All Volatile Water Treatment Yes Building (Section 2.4.2.6)
Page 2-37
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.2-1 Plant Level Scoping Results Description Within Scope of Comments License Renewal?
Auxiliary Building (Section 2.4.2.1) Yes Cable Tunnel (Section 2.4.2.10) Yes Component Supports Commodity Yes Commodity grouping associated with Group (Section 2.4.2.12) Essential Buildings and Yard Structures.
Containment Structures (Section 2.4.1) Yes Control Building (Section 2.4.2.5) Yes Diesel Building (Section 2.4.2.4) Yes Essential Buildings and Yard Yes Structures (Section 2.4.2)
Essential Yard Structures (Section Yes 2.4.2.11)
Intermediate Building (Section Yes 2.4.2.2)
Non-Essential Buildings and Yard No Structures (Section 2.4.3)
Screen House Building (Section Yes 2.4.2.7)
Service Building (Section 2.4.2.9) Yes Standby Auxiliary Feedwater Building Yes (Section 2.4.2.8)
Turbine Building (Section 2.4.2.3) Yes SRP Evaluation Group: Electrical Components 120 VAC Vital Instrument Buses Yes (Section 2.5.2) 125 VDC Power (Section 2.5.3) Yes Page 2-38
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.2-1 Plant Level Scoping Results Description Within Scope of Comments License Renewal?
4160 VAC Power (Section 2.5.4) Yes 480 VAC Power (Section 2.5.5) Yes Control Rod Drive and Nuclear Yes Process Instruments (Section 2.5.6)
Engineered Safety Features Actuation Yes (Section 2.5.10)
Misc. AC Power and Lighting (Section Yes 2.5.7)
Offsite Power (Section 2.5.8) Yes Those portions of the Offsite Power System physically located on site are included within this group.
Plant Communications (Section 2.5.11) Yes Plant Process Computers (Section No 2.5.12)
Plant Security (Section 2.5.13) No Reactor Protection (Section 2.5.9) Yes Those portions of the Circulating Water System that provide circulating water flood detection are evaluated within the Reactor Protection System.
Seismic and Meteorological No Instrumentation (Section 2.5.14)
Page 2-39
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 2.3 Scoping and Screening Results: Mechanical Systems 2.3.1 Reactor Coolant System 2.3.1.1 Reactor Coolant (Class 1)
System Description
NOTE: Reactor Coolant System (RCS) Class 1 components, Steam Generators, the Pressurizer, and the Reactor Vessel are reviewed and evaluated as unique specific topical areas. For clarity purposes the system drawings for Class 1 RCS components include showing the above RCS equipment, with the Class 1 portion clearly denoted with flags.
The Reactor Coolant System transports the heat generated in the reactor core to secondary heat removal systems. The RCS system also acts in conjunction with the fuel and the primary containment system to provide defense in-depth with respect to preventing fission products from escaping to the environment.
Consequently the RCS is associated with mitigating virtually all accidents, transients and events.
The principal components of the Reactor Coolant System include the reactor vessel, pressurizer, steam generators, reactor coolant pumps, and the essential class 1 piping and valves (including the regenerative and letdown heat exchangers). The Reactor Coolant System consists of two identical heat transfer loops connected in parallel to the reactor vessel. Each loop contains a circulating pump and a steam generator.
Aging management reviews for the following system components were performed using the Westinghouse aging management review WCAPs and the corresponding applicant action item requirements detailed in the appropriate NRC Safety Evaluation Report:
- Reactor Vessel
- Reactor Vessel Internals
- Reactor Coolant System Class 1
- Reactor Coolant System Non-Class 1
"* Pressurizer
"* Steam Generator Page 2-40
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information The following fluid systems interface with the Reactor Coolant System:
Plant Sampling Waste Disposal Residual Heat Removal Safety Injection Chemical and Volume Control Component Cooling Water The following RCS subsystems system descriptions are provided below for further detail:
"* Reactor Vessel - Section 2.3.1.2
"* Reactor Vessel Internals - Section 2.3.1.3
"* Pressurizer - Section 2.3.1.4
"* Steam Generators - Section 2.3.1.5 System Function Listing In addition to the System Functions described above, the Reactor Coolant System also contains components which support additional functions (associated system design functions) that may or may not be License Renewal Intended Functions. A comprehensive listing of all functions associated with the system, or specific components contained in the system, is provided in the summary below.
Code A Cri 1 Cri 2 Cri 3 MAINTAIN REACTOR CORE ASSEMBLY GEOMETRY FPIEQ IPTS IATSB Comment: Components within the Reactor Coolant, Core, and Internals system perform this primary design system function. (Reactor Vessel intemals)
Code B Cri 1 Cri 2 Cri 3 INTRODUCE EMERGENCY NEGATIVE REACTIVITY FP EQBPTSBATBSB TO MAKE THE REACTOR SUBCRITICAL x Comment: Components within the Reactor Coolant, Core, and Internals system perform this primary design system function. (Control Rods)
Code D Cri 1 Cri 2 Cri 3 SENSE OR PROVIDE PROCESS CONDITIONS AND FP EQ PTS AT SB GENERATE SIGNALS FOR REACTOR TRIP AND X ENGINEERED SAFETY FEATURES ACTUATION Comment: Components within the Reactor Coolant, Core, and Internals system perform this primary design system function. Reactor Coolant temperature detectors are included as Class 1 boundaries.
Page 2-41
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Code E Cri 1 Cri 2 Cri 3 PROVIDE REACTOR COOLANT PRESSURE FP EQ PTS AT SB BOUNDARY X Comment: Components within the Reactor Coolant, Core, and Internals system perform this primary design system function.
Code F Cri 1 Cri 2 Cri 3 REMOVE RESIDUAL HEAT FROM THE RCS FPEQO PTS AT SB Comment: Components within the Reactor Coolant, Core, and Internals system perform this associated design system function.
Code G Cri 1 Cri 2 Cri 3 PROVIDE EMERGENCY CORE COOLANT WHERE FP EQ PTS ATISB THE ECCS PROVIDES COOLANT DIRECTLY TO THE X CORE Comment: Components within the Reactor Coolant, Core, and Internals system perform this associated design system function.
Code H Cri 1 Cri 2 Cri 3 PROVIDE EMERGENCY HEAT REMOVAL FROM THE FP EQ PTS AT SB REACTOR COOLANT SYSTEM USING SECONDARY X HEAT REMOVAL CAPABILITY I Comment: Components within the Reactor Coolant, Core, and Internals system perform this primary design system function. (Steam Generators)
Code J Cri 1 Cri 2 Cri 3 PROVIDE HEAT REMOVAL FROM SAFETY RELATED FP EQ PTS AT SB HEAT EXCHANGERS X Comment: Components within the Reactor Coolant, Core, and Internals system perform this associated design system function. (Excess Letdown Heat Exchanger and RCP thermal barrier heat exchangers)
Code K Cri 1 Cri 2 Cri 3 PROVIDE PRIMARY CONTAINMENT BOUNDARY FP EQ PTS AT SB Comment: Components within the Reactor Coolant, Core, and Internals system perform this associated design system function.
Code L Cri 1 Cri 2 Cri 3 PROVIDE EMERGENCY HEAT REMOVAL FROM FP FEQ PTS AT SB PRIMARY CONTAINMENT AND PROVIDE X CONTAINMENT PRESSURE CONTROL Comment: Components within the Reactor Coolant, Core, and Internals system perform this associated design system function Page 2-42
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Code R Cri 1 Cri 2 Cri 3 STRUCTURALLY SUPPORT OR HOUSE SAFETY FP EQ PTS AT SB CLASS 1, 2,3 COMPONENTS x Comment: Components within the Reactor Coolant, Core, and Internals system perform this primary design system function. (Reactor Vessel)
Code S Cri 1 Cri 2 Cri 3 SPECIAL CAPABILITY CLASS FUNCTIONS FP IEQ PTS IATI SB Comment: Components within the Reactor Coolant, Core, and Internals system perform this associated design system function (augmented quality). For the purposes of License Renewal, components within the Reactor Coolant, Core, and Internals system that perform special capability class functions are tracked under the Criterion 3 codes (Z1 through Z5). Reg Guide 1.97 Category 3 post accident monitoring variables and mid-loop level detection Code X Cri 1 Cri 2 Cri 3 SFR FUNCTION NOT APPLICABLE AT COMPONENT FPLEQXPTSIATISB LEVEL XI I I I Comment: Components within the Reactor Coolant, Core, and Internals system perform specific safety related functions different from and in addition to the system level functions (e.g. Reg Guide 1.97 Category 1).
Code Z1 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3- SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT X EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR FIRE PROTECTION (10 CFR 50.48)
Comment: Components within the Reactor Coolant, Core, and Internals system perform this function.
Code Z2 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3- SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT x EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR ENVIRONMENTAL QUALIFICATION (10 CFR 50.49)
Comment: Components within the Reactor Coolant, Core, and Internals system perform this function.
Page 2-43
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Code Z3 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3- SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT x EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR PRESSURIZED THERMAL SHOCK (10 CFR 50.61)
Comment: Components within the Reactor Coolant, Core, and Internals system perform this function Code Z5 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3- SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT X EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR STATION BLACKOUT (10 CFR 50.63)
Comment: Components within the Reactor Coolant, Core, and Internals system perform this function.
UFSAR Reference Additional Reactor Coolant System details are provided in Section 5.1, Section 5.2, and Section 5.4 of the UFSAR.
License Renewal Drawings The license renewal drawings for the Reactor Coolant System are listed below:
33013-1246,1 33013-1264 33013-1247 33013-1265,1 33013-1258 33013-1278,1 33013-1260 33013-2248 33013-1262,2 33013-2278 33013-1263 Components Subject to an AMR The component groups for this system that require aging management review are indicated in Table 2.3.1-1 along with each Component Group's passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3.
Page 2-44
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.1-1 Reactor Coolant (Class 1)
Component Group Passive Function Aging Management Reference Class 1 Table 3.2-1 Line Number (1)
Table 3.2-1 Line Number (19)
Table 3.2-1 Line Number (24)
Table 3.2-2 Line Number (13)
VALVES > 4 IN. NPS PRESSURE BOUNDARY Table 3.2-2 Line Number (13)
Table 3.2-2 Line Number (14)
Table 3.2-2 Line Number (34)
Table 3.2-1 Line Number (1)
Table 3.2-1 Line Number (19)
PRESSURE BOUNDARY Table 3.2-2 Line Number (13)
VALVES < 4 IN. NPS Table 3.2-2 Line Number (13)
Table 3.2-2 Line Number (14)
Table 3.2-2 Line Number (34)
Table 3.2-1 Line Number (1)
REACTOR COOLANT (19)
PRESSURE BOUNDARY Table 3.2-1 Line Number PUMPS (CASING AND (24)
Table 3.2-1 Line Number MAIN FLANGE) Table 3.2-2 Line Number (34)
REACTOR COOLANT Table 3.2-2 Line Number (15)
PUMPS (THERMAL PRESSURE BOUNDARY Table 3.2-2 Line Number (34)
BARRIER FLANGE)
THERMAL BARRIER PRESSURE BOUNDARY Table 3.2-2 Line Number (15) This HEAT EXCHANGER TUBING HEAT TRANSFER applies to both passive functions.
Table 3.2-2 Line Number (15) This PRESSURE BOUNDARY applies to both passive functions.
OREUCERSAND ORIFICES ATable 3.2-2 Line Number (34) This REDUCERS THROTTLING applies only to the pressure boundary passive function.
Table 3.2-1 Line Number (1)
PIPING AND FIT-INGS PRESSURE BOUNDARY Table 3.2-1 Line Number (24)
> 4 IN. NPS Table 3.2-2 Line Number (34)
Page 2-45
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.1-1 Reactor Coolant (Class 1)
Component Group Passive Function Aging Management Reference Table 3.2-1 Line Number (1)
PRIMARY LOOP Table 3.2-1 Line Number (10)
P LBOO PRESSURE BOUNDARY Table 3.2-1 Line Number (20)
ELBOWS Table 3.2-2 Line Number (13)
Table 3.2-2 Line Number (34)
Table 3.2-1 Line Number (1)
PIPING AND FITTINGS PRESSURE BOUNDARY Table 3.2-1 Line Number (6)
< 4 IN. NPS Table 3.2-1 Line Number (24)
Table 3.2-2 Line Number (34)
BOLTING FOR FLANGED PIPING MECHANICAL CLOSURE Table 3.2-1 Line Number (22)
JOINTS, RCP'S, AND INTEGRITY Table 3.2-1 Line Number (26)
VALVE CLOSURES REACTOR COOLANT PRESSURE BOUNDARY Table 3.2-2 Line Number (16)
PUMP LUGS Table 3.2-2 Line Number (34) 2.3.1.2 Reactor Vessel Component Description The Ginna Station RPV, as the principal component of the RCS, contains the heat-generating core and associated supports, controls and instrumentation, and coolant circulating channels. Primary outlet and inlet nozzles provide for the exit of heated coolant and its return to the RPV for recirculation through the core.
The Ginna Station RPV consists of a cylindrical shell with a hemispherical bottom head and a flanged and gasketed removable upper head. The RPV shell is fabricated from integral ring forgings joined by circumferential welds.
The RPV contains the core, core support structures, rod control clusters, thermal shield or neutron shield panels, and other parts directly associated with the core. Inlet and outlet nozzles are located at an elevation between the head flange and the core. The body of the RPV is low-alloy carbon steel, and the inside surfaces in contact with coolant are clad with austenitic stainless steel to minimize corrosion. The RPV is supported by steel pads integral with the Page 2-46
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information coolant nozzles. The pads rest on steel base plates atop a support structure attached to the concrete foundation.
UFSAR Reference Additional Reactor Vessel details are provided in Section 5.3 of the UFSAR.
Subcomponents Subject to an Aging Management Review The subcomponents of the Reactor Vessel that require aging management review are indicated in Table 2.3.1-2 along with each subcomponent's passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3.
Table 2.3.1-2 Reactor Vessel Subcomponent Passive Function Aging Management Reference CRDM ROD TRAVEL PRESSURE BOUNDARY Table 3.2-1 Line Number (24)
HOUSINGS Table 3.2-2 Line Number (34)
CRDM LATCH PRESSURE BOUNDARY Table 3.2-1 Line Number (24)
HOUSINGS Table 3.2-2 Line Number (34)
CRDM HOUSING Table 3.2-1 Line Number (23)
TUBES (HEAD PRESSURE BOUNDARY Table 3.2-2 Line Number (2)
ADAPTERS) Table 3.2-2 Line Number (34)
Table 3.2-1 Line Number (23)
VENT PIPE PRESSURE BOUNDARY Table 3.2-2 Line Number (2)
Table 3.2-2 Line Number (34)
Table 3.2-1 Line Number (26)
CLOSURE HEAD PRESSURE BOUNDARY Table 3.2-2 Line Number (1)
DOME Table 3.2-2 Line Number (2)
Table 3.2-2 Line Number (33)
Table 3.2-1 Line Number (26)
CLOSURE HEAD PRESSURE BOUNDARY Table 3.2-2 Line Number (1)
FLANGE Table 3.2-2 Line Number (2)
Table 3.2-2 Line Number (33)
Table 3.2-1 Line Number (26)
PRESSURE BOUNDARY Table 3.2-1 Line Number (28)
VESSEL FLANGE Table 3.2-2 Line Number (1)
SUPPORT RV Table 3.2-2 Line Number (2)
INTERNALS These apply to both passive functions.
Page 2-47
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.1-2 Reactor Vessel Subcomponent Passive Function Aging Management Reference O-RING LEAK PRESSURE BOUNDARY Table 3.2-1 Line Number (9)
MONITOR TUBES Table 3.2-2 Line Number (34)
Table 3.2-1 Line Number (4)
Table 3.2-2 Line Number (1)
Table 3.2-2 Line Number (2)
PRESSURE BOUNDARY Table 3.2-2 Line Number (4)
UPPER SHELL Table 3.2-2 Line Number (4)
Table 3.2-2 Line Number (5)
Table 3.2-2 Line Number (33)
Table 3.2-1 Line Number (4)
Table 3.2-2 Line Number (1)
PRIMARY INLET PRESSURE BOUNDARY Table 3.2-2 Line Number (2)
NOZZLES Table 3.2-2 Line Number (4)
Table 3.2-2 Line Number (5)
Table 3.2-2 Line Number (33)
Table 3.2-1 Line Number (4)
Table 3.2-2 Line Number (1)
PRIMARY OUTLET PRESSURE BOUNDARY Table 3.2-2 Line Number (2)
NOZZLES Table 3.2-2 Line Number (4)
Table 3.2-2 Line Number (5)
Table 3.2-2 Line Number (33)
Table 3.2-1 Line Number (10)
PRIMARY NOZZLE Table 3.2-1 Line Number (24)
SAFE ENDS Table 3.2-2 Line Number (2)
Table 3.2-2 Line Number (34)
Table 3.2-1 Line Number (3)
INTERMEDIATE Table 3.2-1 Line Number (4)
SHELL (INCLUDING PRESSURE BOUNDARY Table 3.2-2 Line Number (1)
CIRCUMFERENTIAL Table 3.2-2 Line Number (2)
WELD) Table 3.2-2 Line Number (5)
Table 3.2-2 Line Number (33)
Table 3.2-1 Line Number (4)
Table 3.2-2 Line Number (1)
LOWER SHELL PRESSURE BOUNDARY Table 3.2-2 Line Number (2)
Table 3.2-2 Line Number (5)
Table 3.2-2 Line Number (33)
Page 2-48
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.1-2 Reactor Vessel Subcomponent Passive Function Aging Management Reference CORE SUPPORT Table 3.2-1 Line Number (9)
SUPPORT RVTal322LieNmr()
LUGS INTERNALS Table 3.2-2 Line Number (2)
Table 3.2-2 Line Number (3)
Table 3.2-2 Line Number (1)
PRESSURE BOUNDARY Table 3.2-2 Line Number (2)
BOTTOM HEAD Table 3.2-2 Line Number (5)
TORUS SUPPORT RV Table 3.2-2 Line Number (33)
INTERNALS These apply to both passive functions.
Table 3.2-2 Line Number (1)
PRESSURE BOUNDARY Table 3.2-2 Line Number (2)
BOTTFOM HEAD DOME Table 3.2-2 Line Number (5)
SUPPORT RV Table 3.2-2 Line Number (33)
INTERNALS These apply to both passive functions.
Table 3.2-1 Line Number (9)
INSTRUMENTATION Table 3.2-2 Line Number (2)
TUBES AND SAFE SUPPORT THIMBLE Table 3.2-2 Line Number (34)
ENDS TUBES These apply to both passive functions.
Table 3.2-1 Line Number (8)
PRESSURE BOUNDARY Table 3.2-2 Line Number (1)
BMI GUIDE TUBES Table 3.2-2 Line Number (2)
SUPPORT THIMBLE Table 3.2-2 Line Number (34)
TUBES These apply to both passive functions.
PRESSURE BOUNDARY Table 3.2-2 Line Number (1)
SEAL TABLE FITTINGS Table 3.2-2 Line Number (34)
SUPPORT THIMBLE These apply to both passive TUBES functions.
VENTILATION Table 3.2-2 Line Number (5)
SHROUD SUPPORT STRUCTURAL SUPPORT Table 3.2-2 Line Number (33)
RING Table 3.2-1 Line Number (18)
CLOSURE STUDS, MECHANICAL CLOSURE Table 3.2-1 Line Number (26)
NUTS, AND WASHERS INTEGRITY Table 3.2-1 Line Number (35)
Table 3.2-2 Line Number (6)
Page 2-49
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.1-2 Reactor Vessel Subcomponent Passive Function Aging Management Reference REFUELING SEAL STRUCTURAL SUPPORT Table 3.2-2 Line Number (5)
LEDGE Table 3.2-2 Line Number (33)
NOZZLE SUPPORT STRUCTURAL SUPPORT Table 3.2-1 Line Number (26)
PADS Table 3.2-2 Line Number (33) 2.3.1.3 Reactor Vessel Internals Component Description The Ginna Station RVIs consist of two basic assemblies, i.e.,
- Upper internals assembly that is removed during each refueling operation to obtain access to the reactor core. The top of this assembly is clamped to a ledge below the vessel-head mating surface by the reactor vessel head. The core barrel fuel alignment pins of the lower internals assembly guides the bottom of the upper internals assembly.
- Lower internals assembly that can be removed, if desired following a complete core unload. This assembly is clamped at the same ledge below the vessel-head mating surface and closely guided at the bottom by radial/clevis assemblies.
UFSAR Reference Additional Reactor Vessel Intemals details are provided in Section 3.9.5 and Section 4.2.1 of the UFSAR.
Subcomponents Subject to an Aging Management Review The subcomponents of the Reactor Vessel Internals that require aging management review are indicated in Table 2.3.1-3 along with each subcomponent's passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3.
Page 2-50
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.1-3 Reactor Vessel Internals Subcomponent Passive Function Aging Management Reference Table 3.2-1 Line Number (8)
Table 3.2-1 Line Number (31)
CORE SUPPORT Table 3.2-1 Line Number (33)
PLATE ANDECORE LOWER COREL PLE Table 3.2-2 Line Number (7)
FLOW DISTRIBUTION Table 3.2-2 Line Number (10)
These apply to both passive functions.
Table 3.2-1 Line Number (8)
Table 3.2-1 Line Number (31)
Table 3.2-1 Line Number (33)
LOWER SUPPORT Table 3.2-2 Line Number (7)
FORGING FLOW DISTRIBUTION Table 3.2-2 Line Number (8)
Table 3.2-2 Line Number (9)
These apply to both passive functions.
Table 3.2-1 Line Number (8)
LOWER SUPPORT CORE SUPPORT Table 3.2-1 Line Number (31)
COLUMNS Table 3.2-1 Line Number (33)
Table 3.2-2 Line Number (7)
Table 3.2-1 Line Number (8)
Table 3.2-1 Line Number (31)
CORE SUPPORT Table 3.2-1 Line Number (33)
CORE BARREL AND LANGEL ATable 3.2-2 Line Number (7)
FLOW DISTRIBUTION Table 3.2-2 Line Number (10)
These apply to both passive functions.
Table 3.2-1 Line Number (8)
RADIAL KEYS AND Table 3.2-1 Line Number (28)
RADIAL KEYS ANDCORE SUPPORT Table 3.2-1 Line Number (33)
CLEVIS INSERTS Table 3.2-2 Line Number (7)
Table 3.2-2 Line Number (9)
Table 3.2-1 Line Number (8)
Table 3.2-1 Line Number (31)
BAFFLE AND FORMER Table 3.2-1 Line Number (33)
ASSEMBLY FLOW DISTRIBUTION Table 3.2-2 Line Number (7)
These apply to both passive functions.
Page 2-51
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.1-3 Reactor Vessel Internals Subcomponent Passive Function Aging Management Reference Table 3.2-1 Line Number (8)
Table 3.2-1 Line Number (31)
CORE BARREL FLOW DISTRIBUTION Table 3.2-1 Line Number (33)
OUTLET NOZZLE Table 3.2-2 Line Number (7)
Table 3.2-2 Line Number (8)
Table 3.2-2 Line Number (9)
SECONDARY CORE CORE SUPPORT Table 3.2-2 Line Number (11) This SUPPORT FLOW DISTRIBUTION applies to both passive functions.
DIFFUSER PLATES FLOW DISTRIBUTION Table 3.2-2 Line Number (11)
Table 3.2-1 Line Number (8)
UPPER SUPPORT GUIDE AND SUPPORT Table 3.2-1 Line Number (33)
PLATE ASSEMBLY RCCA'S Table 3.2-2 Line Number (7)
Table 3.2-2 Line Number (9)
Table 3.2-1 Line Number (8)
Table 3.2-1 Line Number (33)
UPPER CORE PLATE CORE SUPPORT Table 3.2-2 Line Number (7)
AND FUEL Table 3.2-2 Line Number (9)
ALIGNMENT PINS FLOW DISTRIBUTION Table 3.2-2 Line Number (10)
These apply to both passive functions.
Table 3.2-1 Line Number (8)
UPPER SUPPORT GUIDE AND SUPPORT Table 3.2-1 Line Number (33)
COLUMNS RCCA'S Table 3.2-2 Line Number (7)
Table 3.2-2 Line Number (9)
Table 3.2-1 Line Number (8)
RCCA GUIDE TUBES GUIDE AND SUPPORT Table 3.2-1 Line Number (33)
AND FLOW RCCAS Table 3.2-2 Line Number (7)
DOWNCOMERS Table 3.2-2 Line Number (9)
Table 3.2-1 Line Number (8)
GUIDE TUBE GUIDE AND SUPPORT Table 3.2-2 Line Number (7)
SUPPORT PINS RCCAS Table 3.2-2 Line Number (10)
Table 3.2-2 Line Number (11)
Page 2-52
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.1-3 Reactor Vessel Internals Subcomponent Passive Function Aging Management Reference Table 3.2-1 Line Number (8)
Table 3.2-1 Line Number (28)
UPPER CORE PLATE GUIDE AND SUPPORT()
Table 3.2-1 Line Number (7)
Table 3.2-2 Line Number (9)
Table 3.2-1 Line Number (8)
Table 3.2-2 Line Number (3)
Table 3.2-1 Line Number (3)
Table 3.2-2 Line Number (70)
Table 3.2-2 Line Number (92)
Table 3.2-2 Line Number (10)
HOLDDONESPRIN CORE SUPPORT Table 3.2-2 Line Number (11)
Table 3.2-2 Line Number (12)
THERAL SIELDTable 3.2-1 Line Number (8)
HEAD/VESSEL ASCORE SUPPORT Table 3.2-2 Line Number (11)
THERMAL SHIELD Table 3.2-1 Line Number (81)
AND NEUTRON SHIELD VESSEL Tbe321Ln Number ubr(1 Table 3.2-1 Line (33)
PANELS BMI AN OLUMS GUIE AN SUPORT (7) 3.2-1 Line Number (28)
Table 3.2-2 BMI COLUMNS AND GUIDE AND SUPPORT Table 3.2-1 Line Number (28)
FLUX THIMBLES INSTRUMENTATION Table 3.2-2 Line Number (11)
HEAD COOLING SRAY NOOZZLE FLOW DISTRIBUTION Table 3.2-2 Line Number (11)
SPRAY NOZZLES UPPER INSTRUMENTATION GUIDE AND SUPPORT Table 3.2-2 Line Number (11)
COLUMN, CONDUIT THERMOCOUPLES AND SUPPORTS Table 3.2-1 Line Number (8)
Table 3.2-1 Line Number (30)
BOLTING (UPPER Table 3.2-1 Line Number (31)
SUPPORT COLUMN, CORE SUPPORT Table 3.2-1 Line Number (33)
GUIDE TUBE, CLEVIS Table 3.2-1 Line Number (36)
INSERT) Table 3.2-2 Line Number (7)
Table 3.2-2 Line Number (9)
Table 3.2-2 Line Number (12)
Page 2-53
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.1-3 Reactor Vessel Internals Subcomponent Passive Function Aging Management Reference Table 3.2-1 Line Number (5)
Table 3.2-1 Line Number (8)
Table 3.2-1 Line Number (12)
BOLTING SUPPORTI (LOWER COLUM, Table 3.2-1 Line Number (13)
SUPPORT COLUMN, CORE SUPPORT Table 3.2-1 Line Number (31)
BAFFLE/FORMER, Table 3.2-1 Line Number (33)
BARREL/FORMER) Table 3.2-1 Line Number (36)
Table 3.2-2 Line Number (7)
Table 3.2-2 Line Number (12) 2.3.1.4 Pressurizer Component Description The Ginna pressurizer is part of the reactor coolant system (RCS) and is located inside containment. The RCS pressure control consists of the pressurizer vessel equipped with electric heaters, safety valves, relief valves, pressurizer spray, interconnecting piping, and instrumentation. During operation, the pressurizer contains saturated water and steam maintained at the desired saturation temperature and pressure by the electric heaters and pressurizer spray. The chemical and volume control system (CVCS) maintains the desired water level in the pressurizer during steady-state operation by a pressurizer level control instrumentation system.
During normal operation, the external electrical network imposes load changes on the plant turbine generator. These load changes cause temperature changes in the RCS. Since the reactor rod control system which controls the reactor coolant temperature, does not respond instantaneously during a load transient, the pressurizer pressure control system is designed to absorb the reactor coolant volume surges and limit pressure variations during the initial transient period prior to an effective response by the reactor rod control system.
The pressurizer performs the following functions:
"* Maintains the required reactor coolant pressure (pressure boundary function) during steady-state operation and normal heatup and cooldown.
"* Limits pressure changes, to an allowable range, that are caused by reactor coolant thermal expansion and contraction during normal plant load changes and transients.
Page 2-54
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information UFSAR Reference Additional Pressurizer details are provided in Section 5.4.7 of the UFSAR.
Subcomponents Subject to an Aging Management Review The subcomponents of the Pressurizer that require aging management review are indicated in Table 2.3.1-4 along with each subcomponent's passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3.
Table 2.3.1-4 Pressurizer Subcomponent Passive Function Aging Management Reference Table 3.2-1 Line Number (24)
LOWER HEAD PRESSURE BOUNDARY Table 3.2-1 Line Number (26)
Table 3.2-2 Line Number (33)
Table 3.2-1 Line Number (24)
SURGE NOZZLE PRESSURE BOUNDARY Table 3.2-2 Line Number (19)
Table 3.2-2 Line Number (33)
SURGE NOZZLE SAFE PRESSURE BOUNDARY Table 3.2-1 Line Number (24)
END Table 3.2-2 Line Number (34)
HEATER WELL AND PRESSURE BOUNDARY Table 3.2-1 Line Number (24)
HEATER SHEATH Table 3.2-2 Line Number (34)
Table 3.2-1 Line Number (24)
SHELL PRESSURE BOUNDARY Table 3.2-1 Line Number (26)
Table 3.2-2 Line Number (33)
INSTRUMENT Table 3.2-1 Line Number (24)
NOZZLES PRESSURE BOUNDARY Table 3.2-2 Line Number (34)
THERMOWELLS Table 3.2-1 Line Number (24)
UPPER HEAD PRESSURE BOUNDARY Table 3.2-1 Line Number (26)
Table 3.2-2 Line Number (33)
Table 3.2-1 Line Number (24)
SPRAY NOZZLE PRESSURE BOUNDARY Table 3.2-2 Line Number (19)
Table 3.2-2 Line Number (33)
SPRAY NOZZLE SAFE PRESSURE BOUNDARY Table 3.2-1 Line Number (24)
END Table 3.2-2 Line Number (34)
Page 2-55
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.1-4 Pressurizer Subcomponent Passive Function Aging Management Reference Table 3.2-2 Line Number (17)
SAFETY NOZZLE PRESSURE BOUNDARY Table 3.2-2 Line Number (19)
Table 3.2-2 Line Number (33)
SAFETY NOZZLE PRESSURE BOUNDARY Table 3.2-1 Line Number (24)
SAFE END Table 3.2-2 Line Number (34)
Table 3.2-2 Line Number (17)
RELIEF NOZZLE PRESSURE BOUNDARY Table 3.2-2 Line Number (19)
Table 3.2-2 Line Number (33)
RELIEF NOZZLE SAFE PRESSURE BOUNDARY Table 3.2-1 Line Number (24)
END Table 3.2-2 Line Number (34)
Table 3.2-2 Line Number (18)
MANWAY COVER PRESSURE BOUNDARY Table 3.2-2 Line Number (19)
Table 3.2-2 Line Number (33)
SUPPORT SKIRT AND Table 3.2-1 Line Number (26)
SUPPR SSTRUCTURAL SUPPORT Table 3.2-1 Line Number (29)
FLANGE Table 3.2-2 Line Number (33)
Table 3.2-1 Line Number (22)
MECHANICAL CLOSURE Table 3.2-1 Line Number (26)
MANWAY COVER (26)
INTEGRITY Table 3.2-1 Line Number BOLTS Table 3.2-2 Line Number (33) 1 2.3.1.5 Steam Generators Component Description The Steam Generators (SGs) form the boundary between the radioactive primary (Class 1 piping) and the non-radioactive secondary systems. There are two identical steam generators installed in containment, one in each reactor coolant system (RCS) loop. The SG is a vertical shell and tube heat exchanger, where heat transferred from a single-phase fluid at high temperature and pressure (RCS) on the tube side is used to generate a two-phase (steam-water) mixture at a lower temperature and pressure on the shell side.
The reactor coolant flows through the primary side, or inverted U-tubes, entering and leaving through the nozzles located in the hemispherical bottom Page 2-56
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information head of the steam generator. The primary head is divided into inlet and outlet chambers by a vertical partition plate extending from the head to the tube sheet.
The steam-water mixture is generated on the secondary, or shell side.
Feedwater entering the steam generators through a feed ring, mixes with recirculated fluid and flows downward around the tube bundle inner shroud, then enters the tube bundle area where heat is transferred from the RCS. A small portion of the tube bundle located near the tubesheet functions as a preheater to raise the temperature of the fluid to the saturation point. The remaining area of the tube bundle secondary side operates in the heat transfer nucleate boiling region. The wet vapor rises and is dried to a near moisture-free condition as it exits the steam generator at the outlet nozzle at the top of the shell.
At steady-state conditions, the fluid inventory and heat content on both the primary and secondary sides of the steam generator is constant, requiring a virtually constant mass flow on the primary side and a makeup (feedwater) mass flow rate that matches the combined steam flow and blowdown mass flow rates.
UFSAR Reference Additional Steam Generator details are provided in Section 5.4.2 of the UFSAR.
Subcomponents Subject to an Aging Management Review The subcomponents of the Steam Generators that require aging management review are indicated in Table 2.3.1-5 along with each subcomponents' passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3.
Table 2.3.1-5 Steam Generators Subcomponent Passive Function Aging Management Reference Table 3.2-1 Line Number (32)
PRIMARY INLET AND Table 3.2-2 Line Number (20)
OUTLET NOZZLES PRESSURE BOUNDARY Table 3.2-2 Line Number (26)
Table 3.2-2 Line Number (33)
PRIMARY INLET AND Table 3.2-1 Line Number (32)
OUTLET NOZZLE PRESSURE BOUNDARY Table 3.2-2 Line Number (20)
SAFE ENDS Table 3.2-2 Line Number (34)
Page 2-57
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.1-5 Steam Generators Subcomponent Passive Function Aging Management Reference Table 3.2-2 Line Number (21)
TUBESHEET PRESSURE BOUNDARY Table 3.2-2 Line Number (22)
Table 3.2-2 Line Number (22)
DIVIDER PLATE FLOW DISTRIBUTION Table 3.2-2 Line Number (23)
PRESSURE BOUNDARY U-TUBES Table 3.2-1 Line Number (15) This HEAT TRANSFER applies to both passive functions.
Table 3.2-1 Line Number (26)
PRIMARY MANWAYS PRESSURE BOUNDARY Table 3.2-2 Line Number (33)
Table 3.2-2 Line Number (33)
STEAM GENERATOR Table 3.2-1 Line Number (2)
SHELL AND PRESSURE BOUNDARY Table 3.2-2 Line Number (20)
TRANSITION CONE Table 3.2-2 Line Number (33)
Table 3.2-1 Line Number (21)
Table 3.2-2 Line Number (20)
PRESSURE BOUNDARY Table 3.2-2 Line Number (24)
FEEDWATER NOZZLE Table 3.2-2 Line Number (24)
Table 3.2-2 Line Number (33)
Table 3.2-1 Line Number (21)
STEAM OUTLET PRESSURE BOUNDARY Table 3.2-2 Line Number (20)
NOZZLE Table 3.2-2 Line Number (24)
Table 3.2-2 Line Number (33)
STEAM FLOW RESTRICTS FLOW Table 3.2-2 Line Number (24)
RESTRICTOR BLOWDOWN PIPING Table 3.2-1 Line Number (9)
NOZZLES AND Table 3.2-2 Line Number (20)
SECONDARY SIDE PRESSURE BOUNDARY Table 3.2-2 Line Number (20)
SHELL Table 3.2-2 Line Number (33)
PENETRATIONS SECONDARY PRESSURE BOUNDARY Table 3.2-2 Line Number (24)
CLOSURES Table 3.2-2 Line Number (33)
INTERNAL SHROUD, PRIMARY AND STRUCTURAL SUPPORT Table 3.2-2 Line Number (24)
SECONDARY DECKS Page 2-58
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.1-5 Steam Generators Subcomponent Passive Function Aging Management Reference LATTICE GRID TUBE STRUCTURAL SUPPORT Table 3.2-2 Line Number (25)
SUPPORTS U-BEND RESTRAINTS STRUCTURAL SUPPORT Table 3.2-2 Line Number (25)
Table 3.2-1 Line Number (26)
PRIMARY CHANNEL Table 3.2-1 Line Number (32)
HEAD Table 3.2-2 Line Number (20)
Table 3.2-2 Line Number (33)
BOLTSINTERITYTable 3.2-1 Line Number (26)
Table 3.2-1 Line Number (22)
PRIMARY MANWAY MECHANICAL CLOSURE Table 3.2-1 Line Number (26)
EBOLTS INTEGRITY Table 3.2-2 Line Number (33)
SECONDARY SIDE MECHANICAL CLOSURE Table 3.2-1 Line Number (22)
CLOSURE BOLTS INTEGRITY Table 3.2-2 Line Number (33)
Table 3.2-2 Line Number (26)
SUPPORT PADS STRUCTURAL SUPPORT Table 3.2-2 Line Number (27)
Table 3.2-2 Line Number (33)
Table 3.2-2 Line Number (27)
SEISMIC LUGS STRUCTURAL SUPPORT Table 3.2-2 Line Number (33)
Table 3.2-2 Line Number (33) 2.3.1.6 Reactor Coolant (Non-Class 1)
System Description (Non-Class1)
NOTE: Reactor Coolant System (RCS) Class 1 components, Steam Generators, the Pressurizer, and Reactor Vessel are reviewed and evaluated as unique specific topical areas. For clarity purposes the system drawings for Non-Class 1 RCS components include showing the above RCS equipment, but the Class 1 portion is clearly denoted with flags.
The Non-Class 1 Reactor Coolant System (RCS) components system includes all of the safety class 2, 3 and non-nuclear safety grade equipment used to functionally support the Reactor Coolant System. Non-Class 1 RCS equipment is used to sense and provide signals for reactor trip and Engineered Safety Features Actuation. Equipment included within the system boundary is also Page 2-59
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information used for safe shutdown following fires and Station Blackout Events. The Non-Class 1 RCS components system also contains equipment that is Environmentally Qualified.
The principal components of the Non-Class 1 RCS Components system include all RCS interconnected non-class 1 piping instruments and instrument lines, reactor coolant pump motor coolers and heat exchangers, and the pressurizer power operated relief valve (PORV) nitrogen actuation system. Also included within the evaluation boundary are the PORV and safety valve downstream tail piping up to and including the pressurizer relief tank, the reactor vessel level monitoring system, the low RCS loop level instrumentation, in-core nuclear detector drive detector isolation and the essential piping valves and ancillary equipment necessary to support the function of the reactor coolant system.
System Function Listing In addition to the System Functions described above, the Non-Class 1 RCS Components System also contains components which support additional functions (associated system design functions) that may or may not be License Renewal Intended Functions. A comprehensive listing of all functions associated with the system, or specific components contained in the system, is provided in the summary below.
Code D Cri 1 Cri 2 Cri 3 SENSE OR PROVIDE PROCESS CONDITIONS AND FP EQ PTS AT SB GENERATE SIGNALS FOR REACTOR TRIP AND X ENGINEERED SAFETY FEATURES ACTUATION Comment: Components within the Non-Class 1 RCS Components system perform this primary design system function.
Code E Cri 1 Cri 2 Cri 3 PROVIDE REACTOR COOLANT PRESSURE FP EI PTS AT SI BOUNDARY XII Comment: Components within the Non-Class 1 RCS Components system perform this primary design system function.
Code H Cri 1 Cri 2 Cri 3 PROVIDE EMERGENCY HEAT REMOVAL FROM THE FP EQ PTS A REACTOR COOLANT SYSTEM USING SECONDARY X HEAT REMOVAL CAPABILITY Comment: Components within the Non-Class 1 RCS Components system perform this associated design system function. This function is associated with instrument loop power supplies.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Code J Cri 1 Cri 2 Cri 3 PROVIDE HEAT REMOVAL FROM SAFETY RELATED FP EQIPTS AT ISB HEAT EXCHANGERS x Comment: Components within the Non-Class 1 RCS Components system perform this associated design system function. This function is associated with the RCP motor upper bearing oil coolers.
Comment: Components within the Non-Class 1 RCS Components system perform this associated design system function.
Code Q Cri1 Cri2 Cri 3 PROVIDE ELECTRICAL POWER TO SAFETY CLASS FP EQ PTS AT SB 1,2,3 COM PON ENTSx Comment: Components within the Non-Class 1 RCS Components system perform this associated design system function.
Code S Cri 1 Cri 2 Cri 3 SPECIAL CAPABILITY CLASS FUNCTIONS FP EQ PTS AT SB Comment: Components within the Non-Class 1 RCS Components system perform this associated design system function (augmented quality). For the purposes of License Renewal, components within the Non-Class 1 RCS Components system that perform special capability class functions are tracked under the Criterion 2 code (Y) and Criterion 3 codes (Z1 through Z5). The "S"function includes the Digital Metal Impact Monitoring System and Reg Guide 1.97 Category 2 and 3 post accident monitoring variables, non-safety reactor trip signals, etc.
iment: Components within the Non-Class 1 RCS Components system perform this associated design system function.
Components within the Non-Class 1 RCS Components system perform specific safety related functions different from and in addition to the system level functions (e.g. Reg Guide 1.97 Category 1).
Page 2-61
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Code Y Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 2- NON SAFETY FP EQ PTS AT SB RELATED SSC'S WHOSE FAILURE COULD X PREVENT SATISFACTORY ACCOMPLISHMENT OF A SAFETY RELATED FUNCTION Comment: Components within the Non-Class 1 RCS Components system perform this associated design system function. The integrity of the downstream tailpieces from the pressurizer safeties is assumed with respect to the analysis of jet thrust forces during a safety valve lift. The analysis does not extend to the interconnecting pipe which also drains to the pressurizer relief tank Code Z1 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3 - SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT X EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR FIRE PROTECTION (10 CFR 50.48)
Comment: Components within the Reactor Coolant, Core, and Internals system perform this function.
Code Z2 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3- SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT X EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR ENVIRONMENTAL QUALIFICATION (10 CFR 50.49)
Comment: Components within the Reactor Coolant, Core, and Internals system perform this function.
Code Z5 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3- SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT X EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR STATION BLACKOUT (10 CFR 50.63)
Comment: Components within the Non-Class 1 RCS Components system perform this function.
Components Subject to an AMR The component groups for this system that require aging management review are indicated in Table 2.3.1-6 along with each Component Group's passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information UFSAR Reference Additional Reactor Coolant Non-Class 1 details are provided in Section 5.1, Section 5.2, Section 5.4, and Table 6.2-15a of the UFSAR.
License Renewal Drawings The license renewal drawings for the Reactor Coolant Non-Class 1 are listed below:
33013-1246,1 33013-1265,1 33013-1247 33013-1278,1 33013-1258 33013-2248 33013-1260 33013-2278 33013-1262,2 33013-1887 33013-1263 33013-1888 33013-1264 33013-1890 Table 2.3.1-6 Reactor Coolant (Non-Class 1)
Component Group Passive Function Aging Management Reference Table 3.2-2 Line Number (28)
ACCUMULATOR PRESSURE BOUNDARY Table 3.2-2 Line Number (33)
Table 3.2-2 Line Number (33)
CONDENSING PRESSURE BOUNDARY Table 3.2-2 Line Number (28)
CHAMBER Table 3.2-2 Line Number (33)
COOLER PRESSURE BOUNDARY Table 3.2-2 Line Table 3.2-2 Number (31)
Line Number (32)
CS COMPONENTS PRESSURE BOUNDARY Table 3.2-2 Line Number (33)
Table 3.2-2 Line Number (32) 3.2-1 Line Number (26)
Table (33)
INTEGRITY Table 3.2-2 Line Number (BOLTING) Table Table 3.2-2 3.2-2 Line Number Line Number (30)
(33)
PRESSURE BOUNDARY Table 3.2-2 Line Number (31)
HEAT EXCHANGER Table 3.2-2 Line Number (33)
HEAT TRANSFER These apply to both passive functions.
Table 3.2-2 Line Number (28)
OPERATOR PRESSURE BOUNDARY Table 3.2-2 Line Number (33)
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.1-6 Reactor Coolant (Non-Class 1)
Component Group Passive Function Aging Management Reference Table 3.2-1 Line Number (6)
Table 3.2-2 Line Number (29)
PIPE PRESSURE BOUNDARY Table 3.2-2 Line Number (28)
Table 3.2-2 Line Number (29)
Table 3.2-2 Line Number (33)
SUPPORT IN-CORE Table 3.2-2 Line Number (32)
INSTRUMENTATION Table 3.2-2 Line Number (34)
Table 3.2-2 Line Number (28)
PRESSURE BOUNDARY Table 3.2-2 Line Number (34)
STRAINER HOUSING Table 3.2-2 Line Number (34)
TEMPERATURE Table 3.2-1 Line Number (6)
ELEMENT 1 PRESSURE BOUNDARY Table 3.2-2 Line Number (34)
Table 3.2-1 Line Number (6)
VALVE BODY PRESSURE BOUNDARY Tbe322Ln ubr(3 (28)
Table 3.2-2 Line Number Table 3.2-2 Line Number (33)
Table 3.2-2 Line Number (34)
- 1. Selected instruments were conservatively included within the scope of License Renewal. Consideration was given to the consequences of an instrument housing pressure boundary failure. Where an instrument was unisol able from a pressure source and is of sufficient size that a system function would be degraded should the pressure boundary fail, that instrument is included for License Renewal review.
Page 2-64
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 2.3.2 Engineered Safety Features Systems The following systems are addressed in this section:
"* Safety Injection System (Section 2.3.2.1)
"* Containment Spray System (Section 2.3.2.2)
"* Residual Heat Removal System (Section 2.3.2.3)
"* Containment Hydrogen Detectors and Recombiners System (Section 2.3.2.4)
"* Containment Isolation Components (Section 2.3.2.5) 2.3.2.1 Safety Injection (SI)
System Description
The Safety Injection System supports RCS inventory and reactivity control during accident and post-accident circumstances by automatically delivering borated water to the reactor vessel for cooling under high and low reactor coolant pressure conditions. Additionally, the system serves to insert negative reactivity into the Reactor core in the form of borated water during an uncontrolled plant cooldown following a steam line break or an inadvertent valve operation. The Safety Injection System is also credited for use in safe shutdown following some fires and contains components that are part of the Environmental Qualification Program.
Adequate core cooling following a loss-of-coolant accident is provided by the Safety Injection System (SI), which operates as follows:
- 1. Injection of borated water by the passive accumulators.
- 2. Injection by the high-pressure safety injection pumps drawing borated water from the RWST.
- 3. Provide capability for injection by the residual heat removal pumps also drawing borated water from the RWST.
- 4. Recirculation of reactor coolant and injection water from the containment sump to the reactor coolant system by the residual heat removal pumps and the SI pumps, if needed (piggy-back operation).
The principal components of the SI system are two passive accumulators (one for each loop), high-head safety injection pumps, interface with low-head safety injection (Residual Heat Removal pumps), and the essential piping and valves.
The accumulators are passive devices that discharge into the cold leg of each loop. During MODES 1 and 2 the refueling water storage tank (RWST) is aligned to the suction of the high-head safety injection pumps and residual heat removal pumps. The Containment Spray System shares the RWST liquid Page 2-65
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information capacity with the SI system. After the injection phase, coolant spilled from the break and water injected by the safety injection system and the containment spray is cooled and recirculated from the sump to the reactor coolant system by the low-pressure safety injection system or, if needed, by the high-pressure safety injection system.
The following fluid systems interface with the Safety Injection System:
Reactor Coolant Waste Disposal Residual Heat Removal Plant Air Containment Spray Spent Fuel Cooling and Fuel Storage Chemical and Volume Control Component Cooling Water Service Water System Function Listing In addition to the System Functions described above, the Safety Injection System also contains components which support additional functions (associated system design functions) that may or may not be License Renewal Intended Functions. A comprehensive listing of all functions associated with the system, or specific components contained in the system, is provided in the summary below.
Code B Cri 1 Cri 2 Cri 3 INTRODUCE EMERGENCY NEGATIVE REACTIVITY FP EQ PTS AT SB TO MAKE THE REACTOR SUBCRITICAL X Comment: Components within the Safety Injection system perform this primary design system function. The Safety Injection system increases the boron concentration in the Reactor Coolant system during the injection phase of Safety Injection to ensure adequate reactor shutdown margin in the event of a secondary pipe break. The Safety Injection system provides sufficient boron to maintain an adequate post-LOCA sump boron concentration to ensure shutdown of the core. The Safety Injection system delivers borated water to the Reactor Coolant system, as necessary, to compensate for Xenon decay to maintain hot shutdown margin.
Code G Cri 1 Cri 2 Cri 3 PROVIDE EMERGENCY CORE COOLANT WHERE FP EQ PTS AT SB THE ECCS PROVIDES COOLANT DIRECTLY TO THE X CORE Comment: Components within the Safety Injection system perform this primary design system function. The Safety Injection system delivers borated cooling water to the Reactor Coolant system during the injection phase to support core cooling.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Code J Cri 1 Cri 2 Cri 3 PROVIDE HEAT REMOVAL FROM SAFETY RELATED FP EQ PTS AT SB HEAT EXCHANGERS X Comment: Components within the Safety Injection system perform this associated design system function (e.g. seal water heat exchangers and/or heat exchanger interfaces with other fluid systems).
Code K Cri 1 Cri 2 Cri 3 PROVIDE PRIMARY CONTAINMENT BOUNDARY FP IEQ PTS IATSB X
Comment: Components within the Safety Injection system perform this associated design system function. A portion of the Safety Injection system is a Closed Loop Outside Containment (CLOC) pressure boundary.
Code L Cri 1 Cri 2 Cri 3 PROVIDE EMERGENCY HEAT REMOVAL FROM FP EQ PTS ATISB PRIMARY CONTAINMENT AND PROVIDE X CONTAINMENT PRESSURE CONTROL Comment: Components within the Safety Injection system perform this primary design system function. The Safety Injection system provides the liquid capacity of the Refueling Water Storage Tank for the Containment Spray system to provide emergency heat removal from primary containment and provide containment pressure control.
Code S Cri 1 Cri 2 Cri 3 SPECIAL CAPABILITY CLASS FUNCTIONS FP EQ PTS AT SB Comment: Components within the Safety Injection system perform this associated design system function (augmented quality). For the purposes of License Renewal, components within the Safety Injection system that perform special capability class functions are tracked under the Criterion 3 codes (Z1 through Z5). The Safety Injection systems Refueling Water Storage Tank provides a source of borated water during refueling shutdown to flood the refueling cavity.
Additionally, the Safety Injection system provides a backup source of water during normal plant operations for the Chemical and Volume Control system charging pumps in the event the normal source (VCT) is lost. The Refueling Water Storage Tank also provides one of two required boric acid sources sufficient to provide the required shutdown margin at cold shutdown, xenon-free conditions from any expected operating condition.
Page 2-67
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Components within the Safety Injection system perform this associated design system function. The Safety Injection system contains non-nuclear safety class components such as normally isolated sample points and test connections.
Code Y Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 2- NON SAFETY FP EQ PTS AT SB RELATED SSC'S WHOSE FAILURE COULD X PREVENT SATISFACTORY ACCOMPLISHMENT OF A SAFETY RELATED FUNCTION Comment: Components within the Safety Injection system perform this associated design system function. The Safety Injection system contains non-safety piping segments and components whose failure could impact a safety related function.
Code Z1 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3 - SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT X EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR FIRE PROTECTION (10 CFR 50.48)
Comment: Components within the Safety Injection system perform this associated design system function. The safety injection system is capable of isolating RCS boundaries, preventing loss of inventory of the RWST and supplying RWST inventory to the charging system to support reactor coolant make-up capability and reactivity control. Additionally, a Safety Injection pump may be used following some fire events to provide inventory and reactivity control for safe shutdown.
Code Z2 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3- SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT X EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR ENVIRONMENTAL QUALIFICATION (10 CFR 50.49)
Comment: Components within the Safety Injection system are designated as Environmentally Qualified.
UFSAR Reference Additional Safety Injection System details are provided in Section 6.3 and Table 6.2-15a of the UFSAR.
Page 2-68
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information License Renewal Drawings The license renewal drawings for the Safety Injection System are listed below:
33013-1246,2 33013-1262,2 33013-1250,1 33013-1887 33013-1261 33013-1888 33013-1262,1 Components Subject to an AMR The component groups for this system that require aging management review are indicated in Table 2.3.2-1 along with each Component Group's passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3.
Table 2.3.2-1 Safety Injection (SI)
Component Group Passive Function Aging Management Reference Table 3.3-1 Line Number (9)
PRESSURE BOUNDARY 3.3-2 Line Table 3.3-2 Number (1)
ACCUMULATOR Table Line Number (1)
CS COMPONENTS PRESSURE BOUNDARY Table 3.3-1 Line Number (10)
Table 3.3-1 Line Number (10)
Table 3.3-1 Line Number (11)
FASTENERS JOINT INTEGRITY Table 3.3-2 Line Number (9)
(BOLTING) Table 3.3-2 Line Number (10)
Table 3.3-2 Line Number (11)
Table 3.3-1 Line Number (9)
FLOW ELEMENT PRESSURE BOUNDARY Table 3.3-2 Line Number (18)
Table 3.3-2 Line Number (19)
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.2-1 Safety Injection (SI)
Component Group Passive Function Aging Management Reference Table 3.3-1 Line Number (8)
Table 3.3-2 Line Number (23)
Table 3.3-2 Line Number (24)
Table 3.3-2 Line Number (25)
Table 3.3-2 Line Number (30)
Table 3.3-2 Line Number (31)
HEAT TRANSFER These apply to the pressure HEAT EXCHANGER boundary passive function.
PRESSURE BOUNDARY Table 3.3-2 Line Number (26)
Table 3.3-2 Line Number (27)
Table 3.3-2 Line Number (28)
Table 3.3-2 Line Number (29)
These apply to the heat transfer passive function.
INDICATOR' PRESSURE BOUNDARY Table 3.3-1 Line Number (9)
Table 3.3-2 Line Number (34)
Table 3.3-1 Line Number (9) This applies to both passive functions.
PRESSURE BOUNDARY ORIFICE Table 3.3-2 Line Number (38)
RESTRICTS FLOW Table 3.3-2 Line Number (39)
These apply to the pressure boundary passive function.
Table 3.3-1 Line Number (9)
Table 3.3-2 Line Number (46)
Table 3.3-2 Line Number (49)
Table 3.3-2 Line Number (50)
Table 3.3-2 Line Number (54)
PUMP CASING PRESSURE BOUNDARY Table 3.3-2 Line Number (55)
Table 3.3-2 Line Number (56)
Table 3.3-1 Line Number (3)
TANK PRESSURE BOUNDARY Table 3.3-1 Line Number (9)
Table 3.3-2 Line Number (62)
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.2-1 Safety Injection (SI)
Component Group Passive Function Aging Management Reference Table 3.3-1 Line Number (9)
Table 3.3-2 Line Number (71)
Table 3.3-2 Line Number (73)
Table 3.3-2 Line Number (74)
Table 3.3-2 Line Number (76)
Table 3.3-2 Line Number (77)
VALVE BODY PRESSURE BOUNDARY Table 3.3-2 Line Number (78)
Table 3.3-2 Line Number (79)
Table 3.3-2 Line Number (80)
Table 3.3-2 Line Number (81)
Table 3.3-2 Line Number (90)
Table 3.3-2 Line Number (92)
Table 3.3-2 Line Number (93)
- 1. Selected instruments were conservatively included within the scope of License Renewal. Consideration was given to the consequences of an instrument housing pressure boundary failure. Where an instrument was unisol able from a pressure source and is of sufficient size that a system function would be degraded should the pressure boundary fail, that instrument is included for License Renewal review.
Page 2-71
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 2.3.2.2 Containment Spray (CS)
System Description
The Containment Spray (CS) system, in conjunction with the Containment Ventilation system is designed to remove sufficient heat from the containment atmosphere following an accident condition to maintain the containment pressure below design limits. The CS system, in conjunction with the sodium hydroxide (NaOH) tank, is also capable of reducing the iodine and particulate fission product inventories in the containment atmosphere such that the offsite radiation exposure resulting from a LOCA is within the guidelines established by 10 CFR 100. The CS System also contains components that are part of the Environmental Qualification Program.
The principal components of the containment spray system include two pumps, one tank, two spray headers, two eductors, spray nozzles, the essential piping and valves. The system initially takes suction from the refueling water storage tank (RWST). When a low level is reached in the RWST, the spray pump suction is fed from the discharge of the residual heat removal pumps if continued spray is required.
During the period of time that the spray pumps draw from the RWST, approximately 20 gpm of spray additive will be added to the refueling water in each train by using a liquid eductor enabled by the spray pump discharge. The fluid passing from the NaOH tank will then mix with the fluid entering the pump suction. The results will be a solution suitable for the removal of iodine. The containment spray system provides a 100% redundant backup to the containment post-accident charcoal system for iodine removal capability following a LOCA. For operation in the recirculation mode, water is supplied through the residual heat removal pumps.
The following fluid systems interface with the Containment Spray System:
Residual Heat Removal Safety Injection Component Cooling Water Instrument Air Waste Disposal Containment Ventilation Page 2-72
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information System Function Listing In addition to the System Functions listed above, the Containment Spray System also supports additional functions (associated design system functions) that may or may not be License Renewal Intended Functions. A comprehensive listing of all functions associated with the system, or specific components contained in the system, is provided in the summary below.
Code J Cri 1 Cri 2 Cri 3 PROVIDE HEAT REMOVAL FROM SAFETY RELATED FEQPTS AT SB HEAT EXCHANGERS X Comment: Components within the Containment Spray system perform this associated design system function (e.g. seal water heat exchangers cooled by CCW).
Code K Cri 1 Cri 2 Cri 3 PROVIDE PRIMARY ~xs CONTAINMENT BOUNDARY FPFE-Q PTSIATISB Comment: Components within the Containment Spray system perform this associated design system function. A portion of the Containment Spray system outside containment is a closed loop system.
Code L Cri 1 Cri 2 Cri 3 PROVIDE EMERGENCY HEAT REMOVAL FROM FP EQ PTS AT SB PRIMARY CONTAINMENT AND PROVIDE X CONTAINMENT PRESSURE CONTROL Comment: Components within the Containment Spray system perform this primary design system function. The Containment Spray system delivers treated water to the containment spray headers during accident conditions which could over pressurize containment thus ensuring containment pressure does not exceed its design value.
Code M Cri 1 Cri 2 Cri 3 PROVIDE EMERGENCY REMOVAL OF FP EQ PTS AT SB RADIOACTIVE MATERIAL FROM THE PRIMARY X CONTAINMENT ATMOSPHERE Comment: Components within the Containment Spray system perform this primary design system function. The Containment Spray system delivers treated water to the containment spray headers to support removal of elemental iodine from the containment atmosphere in the event of a Loss of Coolant Accident.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Code Q Cri 1 Cri 2 Cri 3 PROVIDE ELECTRICAL POWER TO SAFETY CLASS FPIEQ PTS AT SB 1,2,3 COMPONENTS x Comment: Components within the Containment Spray system perform this associated design system function. (This function denotes protecting the other MCC vital loads from a failure of a load in the Containment Spray system.)
Code S Cri 1 Cri 2 Cri 3 SPECIAL CAPABILITY CLASS FUNCTIONS FP EQ PTS AT SB Comment: Components within the Containment Spray system perform this associated design system function (augmented quality). For the purposes of License Renewal, components within the Containment Spray system that perform special capability class functions are tracked under the Criterion 3 codes (Z1 through Z5).
Code T Cri 1 Cri 2 Cri 3 NON-NUCLEAR SAFETY CLASS FUNCTIONS FP EQ PTS AT SB Comment: Components within the Containment Spray system perform this associated design system function. The Containment Spray system contains non-nuclear safety class components such as normally isolated sample points and test connections.
Code Z2 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3- SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT X EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSIONS REGULATIONS FOR ENVIRONMENTAL QUALIFICATION (10 CFR 50.49)
Comment: Components within the Containment Spray system are designated as Environmentally Qualified.
UFSAR Reference Additional Containment Spray System details are provided in Section 6.5.2, Section 6.2.2, and Table 6.2-15a of the UFSAR.
Page 2-74
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information License Renewal Drawings The license renewal drawings for the Containment Spray System are listed below:
33013-1246,2 33013-1891 33013-1261 Components Subject to an AMR The component groups for this system that require aging management review are indicated in Table 2.3.2-2 along with each Component Group's passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3.
Table 2.3.2-2 Containment Spray (CS)
Component Group Passive Function Aging Management Reference CS COMPONENTS PRESSURE BOUNDARY Table 3.3-1 Line Number (10)
Table 3.3-1 Line Number (9) This PRESSURE BOUNDARY applies to both passive functions.
EDUCTORPROVIDE MIXED FLOW Table 3.3-2 Line Number (8) This applies to the pressure boundary passive function.
Table 3.3-1 Line Number (10)
FASTENERS Table 3.3-1 Line Number (11)
FASTENER JOINT INTEGRITY Table 3.3-2 Line Number (9)
(BOLTING) Table 3.3-2 Line Number (10)
Table 3.3-2 Line Number (11)
Table 3.3-2 Line Number (20)
PRESSUR BOUN Y Table 3.3-2 Line Number (21)
PROVIDE FLOW These apply to both passive functions.
Table 3.3-1 Line Number (8) This HEAT TRANSFER applies to both passive functions.
HEAT EXCHANGER Table 3.3-2 Line Number (23) This applies to the pressure boundary passive function.
INDICATOR 1 PRESSURE BOUNDARY Table 3.3-1 Line Number (9)
Table 3.3-2 Line Number (34)
Page 2-75
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.2-2 Containment Spray (CS)
Component Group Passive Function Aging Management Reference Table 3.3-2 Line Number (35)
PRESSURE BOUNDARY Table 3.3-2 Line Number (36)
ORIFICE Table 3.3-2 Line Number (37)
RESTRICTS FLOW These apply to both passive functions.
Table 3.3-1 Line Number (9)
Table 3.3-2 Line Number (46)
Table 3.3-2 Line Number (49)
PIPETable 3.3-2 Line Number (50)
Table 3.3-2 Line Number (52)
Table 3.3-2 Line Number (53)
Table 3.3-2 Line Number (54)
PUMP CASING PRESSURE BOUNDARY Table 3.3-2 Line Number (55)
Table 3.3-2 Line Number (56)
Table 3.3-2 Line Number (62)
TANK PRESSURE BOUNDARY Table 3.3-2 Line Number (63)
Table 3.3-2 Line Number (64)
TRANSMITTER 1 PRESSURE BOUNDARY Table 3.3-1 Line Number (9)
Table 3.3-2 Line Number (68)
Table 3.3-1 Line Number (9)
Table 3.3-2 Line Number (71)
Table 3.3-2 Line Number (74)
Table 3.3-2 Line Number (76)
Table 3.3-2 Line Number (77)
Table 3.3-2 Line Number (82)
VALVE BODY PRESSURE BOUNDARY Tbe332Ln Table 3.3-2 Line ubr(3 Number (83)
Table 3.3-2 Line Number (90)
Table 3.3-2 Line Number (92)
Table 3.3-2 Line Number (93)
Table 3.3-2 Line Number (95)
Table 3.3-2 Line Number (96)
- 1. Selected instruments were conservatively included within the scope of License Renewal. Consideration was given to the consequences of an instrument housing pressure boundary failure. Where an instrument was unisol able from a pressure source and is of sufficient size that a system function would be degraded should the pressure boundary fail, that instrument is included for License Renewal review.
Page 2-76
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 2.3.2.3 Residual Heat Removal (RHR)
System Description
The Emergency Core Cooling Systems include the use of the Residual Heat Removal System (RHR). The system automatically delivers borated water to the reactor vessel for cooling under low reactor coolant pressure conditions.
The RHR system, in conjunction with the Main and Auxiliary Steam system, is designed to transfer the fission product decay heat and other residual heat from the reactor core to the component cooling water system and the atmosphere at a rate such that design limits of the fuel and the primary system coolant boundary are not exceeded. The RHR system also contains components credited for use in safe shutdown following some fires and components that are part of the Environmental Qualification Program.
Adequate core cooling following a loss-of-coolant accident is provided by the safety injection (emergency core cooling) system, which operates as follows:
- 1. Injection of borated water by the passive accumulators.
- 2. Injection by the high-pressure safety injection pumps drawing borated water from the RWST.
- 3. Injection by the residual heat removal pumps also drawing borated water from the RWST.
- 4. Recirculation of reactor coolant and injection water from the containment sump to the reactor coolant system by the residual heat removal pumps.
The principal components of the RHR system are two RHR (low head safety injection) pumps, two heat exchangers, and the essential piping and valves.
Note: The residual heat removal system discharge line is not used for an Emergency Core Cooling System (ECCS) function that would require MOV-720 or MOV-721 to open; however, a branch of the residual heat removal discharge line provides low-pressure safety injection to the reactor vessel via parallel lines with one normally closed motor-operated valve (MOV-852A or B) and one check valve (CV-853A or B) in each line.
During MODES 1 and 2 the refueling water storage tank (RWST) is aligned to the suction of the high-head safety injection pumps and residual heat removal pumps. After the injection phase, coolant spilled from the break and water injected by the safety injection system and the containment spray is cooled and recirculated to the reactor coolant system by the low-pressure safety injection (residual heat removal) system or, if needed, by the high-pressure safety injection system.
Page 2-77
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information If reactor coolant system depressurization to below the shutoff head of the residual heat removal pumps occurs before the injection mode of the safety injection system is terminated, the residual heat removal pumps will be used in the recirculation mode. The residual heat removal pumps will take suction from the containment sump, circulate the spilled coolant through the residual heat removal heat exchangers, and return the coolant to the reactor via the reactor vessel nozzles. If depressurization of the reactor coolant system proceeds slowly, the high-pressure safety injection pumps are aligned to take suction from the residual heat removal pumps, and inject flow into the reactor coolant system cold legs. The RHR pumps and heat exchangers, in conjunction with the Containment Spray System, may also be used during the recirculation phase to supply water from the containment sump for use in heat and pressure control of the containment atmosphere.
After the steam generators have been used to reduce the reactor coolant temperature to 350 0 F, decay heat cooling is initiated by aligning the residual heat removal pumps to take suction from the reactor coolant system loop A hot leg and discharge through the residual heat removal heat exchangers to the loop B cold leg.
The following fluid systems interface with Residual Heat Removal:
Reactor Coolant Safety Injection Containment Spray Chemical and Volume Control Component Cooling Water System Function Listing In addition to the System Functions described above, the Residual Heat Removal System also contains components which support additional functions (associated design system functions) that may or may not be License Renewal Intended Functions. A comprehensive listing of all functions associated with the system, or specific components contained in the system, is provided in the summary below.
Code F Cri 1 Cri 2 Cri 3
~x HEAT FROM THE RCS REMOVE RESIDUAL FP EQ PTSAT SIB Comment: Components within the Residual Heat Removal system perform this primary design system function. The Residual Heat Removal system, in conjunction with the Safety Injection system, recirculates and cools the water that is collected in the containment sump and returns it to the Reactor Coolant system during the ECCS recirculation phase to support long term cooling.
Page 2-78
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Code G Cri 1 Cri 2 Cri 3 PROVIDE EMERGENCY CORE COOLANT WHERE FP EQ PTS AT SB THE ECCS PROVIDES COOLANT DIRECTLY TO THE X CORE Comment: Components within the Residual Heat Removal system perform this primary design system function. The Residual Heat Removal system, in conjunction with the Safety Injection system, delivers borated cooling water to the Reactor Coolant system during the ECCS injection phase to support core cooling.
Code J Cri 1 Cri 2 Cri 3 PROVIDE HEAT REMOVAL FROM SAFETY RELATED FP EQ PTS AT SB HEAT EXCHANGERS X Comment: Components within the Residual Heat Removal system perform this associated design system function (e.g. seal water heat exchangers).
Components within the Residual Heat Removal system perform this associated design system function. The Residual Heat Removal system has containment isolation valves. A portion of the Residual Heat Removal system outside containment is a closed loop system.
Code L Cri 1 Cri 2 Cri 3 PROVIDE EMERGENCY HEAT REMOVAL FROM FP EQ PTS AT SB PRIMARY CONTAINMENT AND PROVIDE X CONTAINMENT PRESSURE CONTROL Comment: Components within the Residual Heat Removal system perform this primary design system function. The Residual Heat Removal system provides the capability to supply water to the suction of the Containment Spray pumps when in recirculation mode.
Code S Cri 1 Cri 2 Cri 3 SPECIAL CAPABILITY CLASS FUNCTIONS FP FEQ PTS AT SB Comment: Components within the Residual Heat Removal system perform this associated design system function (augmented quality). For the purposes of License Renewal, components within the Residual Heat Removal system that perform special capability class functions are tracked under the Criterion 3 codes (Z1 through Z5). The Residual Heat Removal system removes sensible and decay heat from the Reactor Coolant system (RCS) during cooldown, cold shutdown and refueling shutdown.
Page 2-79
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Code T Cri 1 Cri 2 Cri 3 NON-NUCLEAR SAFETY CLASS FUNCTIONS FPE PTS IATISB Comment: Components within the Residual Heat Removal system perform this associated design system function. The Residual Heat Removal system contains non-nuclear safety class components such as normally isolated sample points and test connections.
Code Z1 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3 - SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT x EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR FIRE PROTECTION (10 CFR 50.48)
Comment: Components within the Residual Heat Removal system perform this associated design system function. The Residual Heat Removal system may be used following some fire events to provide reactivity control and decay heat removal for safe shutdown.
Code Z2 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3- SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT X EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR ENVIRONMENTAL QUALIFICATION (10 CFR 50.49)
Comment: Components within the Residual Heat Removal system are designated as Environmentally Qualified.
UFSAR Reference Additional Residual Heat Removal System details are provided in Section 6.3.2.3, Section 5.4.5, and Table 6.2-1 5a of the UFSAR.
Page 2-80
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information License Renewal Drawings The license renewal drawings for the Residual Heat Removal System are listed below:
33013-1245 33013-1272,2 33013-1247 33013-1278,2 33013-1260 33013-1890 33013-1264 Components Subject to an AMR The component groups for this system that require aging management review are indicated in Table 2.3.2-3 along with each Component Group's passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3.
Table 2.3.2-3 Residual Heat Removal (RHR)
Component Group Passive Function Aging Management Reference CS COMPONENTS PRESSURE BOUNDARY Table 3.3-1 Line Number (10)
Table 3.3-1 Line Number (10)
FASTENERS Table 3.3-1 Line Number (11)
FASTENER JOINT INTEGRITY Table 3.3-2 Line Number (9)
(BOLTING) Table 3.3-2 Line Number (10)
Table 3.3-2 Line Number (11)
Table 3.3-1 Line Number (9)
FLOW ELEMENT PRESSURE BOUNDARY Table 3.3-2 Line Number (9)
Table 3.3-2 Line Number (19)
Table 3.3-1 Line Number (8) This applies to both passive functions.
SHEAT TRANSFER Table 3.3-2 Line Number (22)
HEAT EXCHANGER Table 3.3-2 Line Number (23)
Table 3.3-2 Line Number (33)
These apply to the pressure boundary passive function.
INDICATOR' PRESSURE BOUNDARY Table 3.3-1 Line Number (9)
Table 3.3-2 Line Number (34)
Page 2-81
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.2-3 Residual Heat Removal (RHR)
Component Group Passive Function Aging Management Reference Table 3.3-1 Line Number (9) This PRESSURE BOUNDARY applies to both passive functions.
ORIFICERESTRICTS FLOW Table 3.3-2 Line Number (39) This applies to the pressure boundary passive function.
Table 3.3-1 Line Number (9)
Table 3.3-2 Line Number (46)
PIPE PRESSURE BOUNDARY Table 3.3-2 Line Number (48)
Table 3.3-2 Line Number (49)
Table 3.3-2 Line Number (50)
Table 3.3-1 Line Number (9)
PUMP CASING PRESSURE BOUNDARY Table 3.3-2 Line Number (5)
Table 3.3-2 Line Number (58)
SWITCH' PRESSURE BOUNDARY Table 3.3-1 Line Number (9)
Table 3.3-2 Line Number (61)
TEMPERATURE Table 3.3-1 Line Number (9)
ELEMENT 1 PRESSURE BOUNDARY Table 3.3-2 Line Number (65)
Table 3.3-1 Line Number (9)
Table 3.3-2 Line Number (71)
Table 3.3-2 Line Number (73)
VALVE BODY PRESSURE BOUNDARY Table 3.3-2 Line Tbe332Ln Number ubr(6 (74)
Table 3.3-2 Line Number (76)
Table 3.3-2 Line Number (77)
Table 3.3-2 Line Number (92)
Table 3.3-2 Line Number (93)
I. Selected instruments were conservatively included within the scope of License Renewal. Consideration was given to the consequences of an instrument housing pressure boundary failure. Where an instrument was unisol able from a pressure source and is of sufficient size that a system function would be degraded should the pressure boundary fail, that instrument is included for License Renewal review.
Page 2-82
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 2.3.2.4 Containment Hydrogen Detectors and Recombiners
System Description
Two trains of containment hydrogen detectors and hydrogen recombiner units are available to the plant. The purpose of these units is to detect and control combustible gas mixtures in the primary containment atmosphere. Portions of these trains are Environmentally Qualified. Because containment hydrogen buildup is a relatively slow process, the recombiner equipment located outside of containment is maintained at a lesser degree of prompt readiness than any other engineered safety feature. Those portions of the recombiner system are considered non-safety related components whose failure could prevent the satisfactory accomplishment of a safety related function.
The principal components of the Containment Hydrogen Detection and Recombiner System include:
For the detection portion, two hydrogen concentration monitoring devices, a local analyzer/control panels, remote monitoring/control panels and their corresponding essential piping and valves. The hydrogen monitoring system is capable of operation during post-accident conditions. The monitors are normally maintained in an isolated standby mode. The recombiner portion consists of two blowers and combustion chambers complete with main burner, two igniters (one a spare), pilot burner, and a dilution chamber, two control panels and the corresponding essential piping and valves. Each combustor is fired by an externally supplied fuel gas, employing containment air as the oxidant. The air supply blowers deliver primary combustion air and quench air to reduce the unit exhaust temperature. Hydrogen in the containment air is oxidized in passing through the combustion chamber. Hydrogen gas is also used as the externally supplied fuel so that noncondensible combustion products, which would cause a progressive rise in containment pressure, are avoided. Oxygen gas is made up through a separate containment feed to prevent depletion of containment oxygen below the concentration required for stable operation of the combustor.
The following fluid systems interface with the Containment Hydrogen Detection and Recombiner System:
Waste Disposal Plant Air Page 2-83
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information System Function Listing In addition to the System Functions described above, the Containment Hydrogen Detection and Recombiner System also contains components which support additional functions (associated system design functions) that may or may not be License Renewal Intended Functions. A comprehensive listing of all functions associated with the system, or specific components contained in the system, is provided in the summary below.
Comment: Components within the Containment Hydrogen Detectors and Recombiners system perform this associated design system function.
Code N Cri 1 Cri 2 Cri 3 CONTROL COMBUSTIBLE GAS MIXTURES INTHE FP EQ PTS ATISB PRIMARY CONTAINMENT ATMOSPHERE X Comment: Components within the Containment Hydrogen Detectors and Recombiners system perform this primary design system function.
Code 0 Cri 1 Cri 2 Cri 3 PROVIDE ELECTRICAL POWER TO SAFETY CLASS FP EQ PTS AT SB 1,2,3 COMPONENTS XI Comment: Components within the Containment Hydrogen Detectors and Recombiners system perform this associated design system function.
Code S Cri 1 Cri 2 Cri 3 SPECIAL CAPABILITY CLASS FUNCTIONS EP FEQ PTS AT SB Comment: Components within the Containment Hydrogen Detectors and Recombiners system perform this associated design system function (augmented quality). For the purposes of License Renewal, components within the Containment Hydrogen Detectors and Recombiners system that perform special capability class functions are tracked under the System Function code Y (Criterion 2).
Comment: Components within the Containment Hydrogen Detectors and Recombiners system perform this associated design system function.
Page 2-84
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Code X Cri 1 Cri 2 Cri 3 SFR FUNCTION NOT APPLICABLE AT COMPONENT FP EQ PTS A LEVEL X Comment: Components within the Containment Hydrogen Detectors and Recombiners system perform specific safety related functions different from and in addition to the system level functions (e.g. Reg Guide 1.97 Category 1).
Code Y Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 2- NON SAFETY FP EQ PTS AT SB RELATED SSC'S WHOSE FAILURE COULD X PREVENT SATISFACTORY ACCOMPLISHMENT OF A SAFETY RELATED FUNCTION Comment: Components within the Containment Hydrogen Detectors and Recombiners system perform this associated design system function.
Some Containment Hydrogen Detectors and Recombiners components located outside containment are maintained in a quality program but are not classified as nuclear safety related. The basis for this determination is that this standby system has ample time for operability testing and repairs prior to being placed in service. Additionally the system requires the delivery of vendor supplied gases for operation. This not-withstanding, those portions of the system which control or transport the gas mixtures outside of containment are conservatively categorized for License Renewal purposes as SSC's whose failure could prevent the satisfactory accomplishment of a safety related function.
Code Z2 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3- SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT X EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR ENVIRONMENTAL QUALIFICATION (10 CFR 50.49)
Comment: Components within the Containment Hydrogen Detectors and Recombiners system are designated as Environmentally Qualified.
UFSAR Reference Additional Containment Hydrogen Detection and Recombiner System details are provided in Section 6.2.5, Section 1.5.10, and Table 6.2-15a of the UFSAR.
Page 2-85
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information License Renewal Drawings The license renewal drawings for the Containment Hydrogen Detection and Recombiner System are listed below:
33013-1274 33013-1889 33013-1275.1 33013-1892 33013-1275,2 33013-1899,1 33013-1278,1 Components Subject to an AMR The component groups for this system that require aging management review are indicated in Table 2.3.2-4 along with each Component Group's passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3.
Table 2.3.2-4 Containment Hydrogen Detectors and Recombiners Component Group Passive Function Aging Management Reference Table 3.3-2 Line Number (2)
BLOWER CASING PRESSURE BOUNDARY Table 3.3-2 Line Number (3)
Table 3.3-2 Line Number (3)
Table 3.3-2 Line Number (4)
CONTROLLER 1 PRESSURE BOUNDARY Table 3.3-2 Line Number (5)
Table 3.3-2 Line Number (6)
CS COMPONENTS PRESSURE BOUNDARY Table 3.3-1 Line Number (10)
Table 3.3-1 Line Number (10)
FASTENERS Table 3.3-2 Line Number (11)
FASTENER JOINT INTEGRITY Table 3.3-2 Line Number (9)
(BOLTING) Table 3.3-2 Line Number (10)
Table 3.3-2 Line Number (11)
Table 3.3-2 Line Number (12)
FILTER HOUSING PRESSURE BOUNDARY Table 3.3-2 Line Number (13)
Table 3.3-2 Line Number (13)
FLOW ELEMENT PRESSURE BOUNDARY Table 3.3-2 Line Number (17)
Table 3.3-2 Line Number (17)
Page 2-86
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.2-4 Containment Hydrogen Detectors and Recombiners Component Group Passive Function Aging Management Reference Table 3.3-1 Line Number (3)
Table 3.3-2 Line Number (40)
Table 3.3-2 Line Number (42)
Table 3.3-2 Line Number (43)
Table 3.3-2 Line Number (44)
PIPETable 3.3-2 Line Number (45)
Table 3.3-2 Line Number (46)
Table 3.3-2 Line Number (49)
Table 3.3-2 Line Number (50)
Table 3.3-2 Line Number (51)
PUMP CASING PRESSURE BOUNDARY Table 3.3-2 Line Number (57)
Table 3.3-2 Line Number (58)
RECOMBINER PRESSURE BOUNDARY Table 3.3-2 Line Number (59)
CASING Table 3.3-2 Line Number (60)
Table 3.3-1 Line Number (3)
Table 3.3-2 Line Number (69)
Table 3.3-2 Line Number (71)
Table 3.3-2 Line Number (74)
Table 3.3-2 Line Number (84)
Table 3.3-2 Line Number (85)
Table 3.3-2 Line Number (86)
Table 3.3-2 Line Number (88)
Table 3.3-2 Line Number (89)
Table 3.3-2 Line Number (90)
Table 3.3-2 Line Number (91)
Table 3.3-2 Line Number (92)
Table 3.3-2 Line Number (93)
Table 3.3-2 Line Number (94)
VENTILATION PRESSURE BOUNDARY Table 3.3-2 Line Number (97)
DUCTWORK Table 3.3-2 Line Number (98)
- 1. Selected instruments were conservatively included within the scope of License Renewal. Consideration was given to the consequences of an instrument housing pressure boundary failure. Where an instrument was unisol able from a pressure source and is of sufficient size that a system function would be degraded should the pressure boundary fail, that instrument is included for License Renewal review.
Page 2-87
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 2.3.2.5 Containment Isolation Components
System Description
Some plant support systems have no license renewal intended functions at the system level as described in the Updated Final Safety Analysis Report (UFSAR) but do have piping segments and valves that act as part of the containment pressure boundary. The Containment structure itself has piping and valves used for containment pressure sensors and for provisions to test the containment access hatches and containment isolation flange o-rings. The Containment Isolation Components System contains the non-structural equipment detailed in the UFSAR as performing a containment isolation boundary function where the system containing that equipment has no other safety related system function. Components evaluated in the Containment Isolation Components system are relied upon to achieve safe shutdown following some fires and contains components that are part of the Environmental Qualification Program.
The principal components of the Containment Isolation Boundary Components System include pipes and valves. A summary of the system lines penetrating Containment and the boundaries employed for containment isolation is presented in UFSAR Table 6.2-15a.
Each system whose piping penetrates the Containment boundary is designed to maintain or establish isolation of the Containment from the outside environment under any accident for which isolation is required, and assuming a coincident independent single failure or malfunction occurring in any active system component within the isolated bounds. Piping penetrating the Containment is designed for pressures at least equal to the containment design pressure. Containment isolation boundaries are provided as necessary in lines penetrating the Containment to ensure that no unrestricted release of radioactivity can occur.
The following Mechanical Systems interface with the Containment Isolation Components System:
Plant Air Systems Plant Sampling Heating Steam Page 2-88
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information System Function Listing In addition to the System Functions described above, the Containment Isolation Boundary Components System also contains components which support additional functions (associated system design functions) that may or may not be License Renewal Intended Functions. A comprehensive listing of all functions associated with the system, or specific components contained in the system, is provided in the summary below.
Code K Cri 1 Cri 2 Cri 3 PROVIDE PRIMARY CONTAINMENT BOUNDARY FP EQ PTS AT SIB x
Comment: Components within the Containment Isolation Components system perform this primary design system function. Some plant support systems have no license renewal intended functions at the system level as described in the Updated Final Safety Analysis Report (UFSAR) but do have piping segments and valves that act as part of the containment pressure boundary. The Containment structure itself has piping and valves used for containment pressure sensors and for provisions to test the containment access hatches and containment isolation flange o-rings.
Code S Cri 1 Cri 2 Cri 3 SPECIAL CAPABILITY CLASS FUNCTIONS FP EQ PTS AT SB Comment: Components within the Containment Isolation Components system perform this associated design system function (augmented quality). For the purposes of License Renewal, components within the Containment Isolation Components system that perform License Renewal Criterion 3 functions are tracked under the Criterion 3 codes (Z1 through Z5).
Components within the Containment Isolation Components system support monitoring Reg. Guide 1.97 Cat 3 variables.
Code Z1 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3- SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT x EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR FIRE PROTECTION (10 CFR 50.48) i Comment: Components within the Containment Isolation Components system perform this associated design system function. Primary sample containment isolation valves are credited with maintaining RCS inventory for safe shutdown following fires.
Page 2-89
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Code Z2 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3 - SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT x EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR ENVIRONMENTAL QUALIFICATION (10 CFR 50.49)
Comment: Components within the Containment Isolation Components system are designated as Environmentally Qualified (RCS sample containment isolation solenoid valve 14104S).
UFSAR Reference Additional Containment Isolation Components System details are provided in Section 6.2.4 and Table 6.2-15a of the UFSAR.
License Renewal Drawings The license renewal drawings for the Containment Isolation Components System are listed below:
33013-1248 33013-1886,2 33013-1261 33013-1893 33013-1278,1 33013-1915 33013-1279 33013-1887 33013-1882 33013-1888 33013-1884,1 33013-1890 33013-1884,2 33013-1899,1 Components Subject to an AMR The component groups for this system that require aging management review are indicated in Table 2.3.2-5 along with each Component Group's passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3.
Table 2.3.2-5 Containment Isolation Components Component Group Passive Function Aging Management Reference CS COMPONENTS PRESSURE BOUNDARY Table 3.3-1 Line Number (10)
Table 3.3-1 Line Number (9)
DELAY COIL PRESSURE BOUNDARY Table 3.3-2 Line Number (7) 1 Table 3.3-2 Line Number (7)
Page 2-90
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.2-5 Containment Isolation Components Component Group Passive Function Aging Management Reference Table 3.3-1 Line Number (10)
FASTENERS Table 3.3-2 Line Number (11)
FASTENER JOINT INTEGRITY Table 3.3-2 Line Number (9)
(BOLTING) Table 3.3-2 Line Number (10)
Table 3.3-2 Line Number (11)
Table 3.3-1 Line Number (4)
FLANGE PRESSURE BOUNDARY Table 3.3-2 Line Number (14)
Table 3.3-2 Line Number (15)
Table 3.3-1 Line Number (4)
Table 3.3-1 Line Number (9)
Table 3.3-2 Line Number (40)
Table 3.3-2 Line Number (41)
PIPE PRESSURE BOUNDARY Table 3.3-2 Line Number (43)
Table 3.3-2 Line Number (44)
Table 3.3-2 Line Number (47)
Table 3.3-2 Line Number (49)
Table 3.3-2 Line Number (50)
Table 3.3-2 Line Number (66)
THERMOWELL PRESSURE BOUNDARY Table 3.3-2 Line Number (67)
Table 3.3-2 Line Number (67)
Table 3.3-1 Line Number (4)
Table 3.3-1 Line Number (9)
Table 3.3-2 Line Number (70)
Table 3.3-2 Line Number (72)
Table 3.3-2 Line Number (73)
Table 3.3-2 Line Number (74)
PRESSURE BOUNDARY Table 3.3-2 Line Number (75)
VALVE BODY Table 3.3-2 Line Number (86)
Table 3.3-2 Line Number (87)
Table 3.3-2 Line Number (88)
Table 3.3-2 Line Number (89)
Table 3.3-2 Line Number (91)
Table 3.3-2 Line Number (92)
Table 3.3-2 Line Number (93)
Page 2-91
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 2.3.3 Auxiliary Systems The following systems are addressed in this section:
"* Chemical Volume and Control System (Section 2.3.3.1)
"* Component Cooling Water System (Section 2.3.3.2)
"* Spent Fuel Cooling and Fuel Storage System (Section 2.3.3.3)
"* Waste Disposal System (Section 2.3.3.4)
"* Service Water System (Section 2.3.3.5)
"* Fire Protection System (Section 2.3.3.6)
"* Heating Steam (Section 2.3.3.7)
"* Emergency Power System (Section 2.3.3.8)
"* Containment Ventilation Systems (Section 2.3.3.9)
"* Essential Ventilation Systems (Section 2.3.3.10)
"* Cranes, Hoists, and Lifting Devices (Section 2.3.3.11)
"* Treated Water System (Section 2.3.3.12)
"* Radiation Monitoring (Section 2.3.3.13)
"* Circulating Water - Not Within Scope of License Renewal (Section 2.3.3.14)
"* Chilled Water - Not Within Scope of License Renewal (Section 2.3.3.15)
"* Fuel Handling - Not Within Scope of License Renewal (Section 2.3.3.16)
"* Plant Sampling - Not Within Scope of License Renewal (Section 2.3.3.17)
"* Plant Air - Not Within Scope of License Renewal (Section 2.3.3.18)
"* Non-Essential Ventilation - Not Within Scope of License Renewal (Section 2.3.3.19)
"* Site Service and Facility Support - Not Within Scope of License Renewal (Section 2.3.3.20) 2.3.3.1 Chemical and Volume Control (CVCS)
System Description
The Chemical and Volume Control System (CVCS) controls and maintains reactor coolant system inventory and purity through the process of makeup and letdown, and provides seal injection flow to the reactor coolant pump seals. In addition to the reactivity control achieved by the control rods, reactivity control is provided by CVCS, which regulates the concentration of boric acid solution neutron absorber in the reactor coolant system. In order to perform the above Page 2-92
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information functions a continuous feed-and-bleed is maintained between the reactor coolant system and the chemical and volume control system. The CVCS is also credited for use in safe shutdown following Station Blackout events and some fire events. Selected large volume CVCS tanks are considered non safety equipment whose failure could affect a safety function due to their potential to cause flooding effects.
The principal components of CVCS are variable speed charging pumps, tanks, heat exchangers, demineralizers, and the essential piping and valves. The chemical and volume control system controls and maintains reactor coolant system inventory and purity through the process of makeup and letdown, and provides seal injection flow to the reactor coolant pump seals. The letdown portion of the system consists of a regenerative heat exchanger and a nonregenerative heat exchanger to cool the reactor coolant letdown and three parallel orifice valves to reduce the pressure. The coolant is passed through purification and deborating demineralizers, as necessary, where corrosion and fission products are removed. The coolant is then routed to the volume control tank. Seal return flow passes from the reactor coolant pump seals, through a containment isolation valve and the seal-water heat exchanger, before returning to the volume control tank. The seal return line is at low pressure and temperature. The charging pumps draw from the volume control tank and inject into the reactor coolant system, both through the normal makeup path and via the reactor coolant pump seals.
The following fluid systems interface with the Chemical and Volume Control System:
Reactor Coolant Waste Disposal Residual Heat Removal Instrument Air Spent Fuel Cooling and Fuel Storage Service Water Component Cooling Water Treated Water Page 2-93
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information System Function Listing In addition to the System Functions described above, the CVCS also contains components which support additional functions (associated system design functions) that may or may not be License Renewal Intended Functions. A comprehensive listing of all functions associated with the system, or specific components contained in the system, is provided in the summary below.
Code J On 1 Cri2 Cri 3 PROVIDE HEAT REMOVAL FROM SAFETY RELATD FPIEQIPTSI ATISB HEAT EXCHANGERS XD' Comment: Components within the Chemical and Volume Control system perform this associated design system function.
Code K Cri 1 Cri 2 Cri 3 PROVIDE PRIMARY CONTAINMENT BOUNDARY FP EQ PTS AT SB x
Comment: Components within the Chemical and Volume Control system perform this associated design system function.
Code L Cri 1 Cri 2 Cri 3 PROVIDE EMERGENCY HEAT REMOVAL FROM FP EQ PTS AT SB PRIMARY CONTAINMENT AND PROVIDE X CONTAINMENT PRESSURE CONTROL Comment: Components within the Chemical and Volume Control system perform this associated design system function. (Interface with RWST)
Code Q Cri 1 Cri 2 Cri 3 PROVIDE ELECTRICAL POWER TO SAFETY CLASS FPIEQtPTSIAT SB 1,2,3 COM PONE NTS xII II Comment: Components within the Chemical and Volume Control system perform this associated design system function.
Code S Cri 1 Cri 2 Cri 3 SPECIAL CAPABILITY CLASS FUNCTIONS FP EQ PTS AT SB Comment: Components within the Chemical and Volume Control system perform this associated design system function (augmented quality). For the purposes of License Renewal, components within the Chemical and Volume Control system that perform special capability class functions are tracked under the Criterion 3 codes (Z1 through Z5). The "S"function also includes identifying components which provide or maintain sufficient reactor coolant inventory and reactivity control to achieve and maintain normal shutdown conditions.
Page 2-94
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Code T Cri 1 Cri 2 Cri 3 NON-NUCLEAR SAFETY CLASS FUNCTIONS FP EQ PTS AT SB Comment: Components within the Chemical and Volume Control system perform this associated design system function.
Code X Cri 1 Cri 2 Cri 3 SFR FUNCTION NOT APPLICABLE AT COMPONENT VFP EQ PTS AT SB LEVEL X Comment: Components within the Chemical and Volume Control system perform specific safety related functions different from and in addition to the system level functions. (AOV-392A performs both a relief and an isolation function.)
Code Y Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 2- NON SAFETY FP EQ PTS AT SB RELATED SSC'S WHOSE FAILURE COULD X PREVENT SATISFACTORY ACCOMPLISHMENT OF A SAFETY RELATED FUNCTION I Comment: Components within the Chemical and Volume Control system perform this associated design system function. This function identifies non-safety tanks whose failure can affect safety related systems due to flooding.
(BAST's, CVCS HUT's, RMW Tank, and VCT). In addition, some non-safety pipe is used to provide a relief path protecting containment isolation valves and piping.
Code Zi Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3 - SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT X EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR FIRE PROTECTION (10 CFR 50.48)
Comment: Components within the Chemical and Volume Control system perform this associated design system function.
Code Z5 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3- SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT X EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR STATION BLACKOUT (10 CFR 50.63)
Comment: Components within the Chemical and Volume Control system perform this associated design system function.
Page 2-95
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information UFSAR Reference Additional Chemical and Volume Control System details are provided in Section 9.3.4 and Table 6.2-15a of the UFSAR.
License Renewal Drawings The license renewal drawings for the Chemical and Volume Control System are listed below:
33013-1245 33013-1269 33013-1246,2 33013-1274 33013-1247 33013-1278,2 33013-1258 33013-1891 33013-1260 33013-1887 33013-1262,1 33013-1888 33013-1264 33013-1889 33013-1265,1 33013-1890 33013-1265,2 33013-1892 33013-1266 33013-2274 33013-1267 33013-2275,1 33013-1268 33013-2275,2 Components Subject to an AMR The component groups for this system that require aging management review are indicated in Table 2.3.3-1 along with each Component Group's passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3.
Table 2.3.3-1 Chemical and Volume Control (CVCS)
Component Group Passive Function Aging Management Reference Table 3.4-2 Line Number (11)
Table 3.4-2 Line Number (12)
CONDENSER PRESSURE BOUNDARY Table 3.4-2 Line Tbe342Ln Number ubr(4 (13)
Table 3.4-2 Line Number (14)
Table 3.4-2 Line Number (15)
Table 3.4-2 Line Number (16)
Table 3.4-2 Line Number (27)
Table 3.4-2 Line Number (28)
PRESSURE BOUNDARY Table 3.4-2 Line Number (29)
COOLER Table 3.4-2 Line Number (30)
Table 3.4-2 Line Number (31)
Table 3.4-2 Line Number (32)
Page 2-96
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-1 Chemical and Volume Control (CVCS)
Component Group Passive Function Aging Management Reference CS COMPONENTS PRESSURE BOUNDARY Table 3.4-1 Line Number (13)
Table 3.4-1 Line Number (13)
FASTENERS Table 3.4-1 Line Number (23)
FASTENES JOINT INTEGRITY Table 3.4-2 Line Number (79)
(BOLTING) Table 3.4-2 Line Number (80)
Table 3.4-2 Line Number (81)
Table 3.4-2 Line Number (99)
FILTER HOUSING PRESSURE BOUNDARY Table 3.4-2 Line Number (101)
Table 3.4-2 Line Number (102)
Table 3.4-2 Line Number (107)
Table 3.4-2 Line Number (108)
PRESSURE BOUNDARY Table 3.4-2 Line Number (108)
FLOW ELEMENT Table 3.4-2 Line Number (109)
Table 3.4-2 Line Number (110)
Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (8)
HEAT EXCHANGER PRESSURE BOUNDARY Table 3.4-1 Line Number (14)
Table 3.4-2 Line Number (144)
Table 3.4-2 Line Number (151)
Table 3.4-2 Line Number (235)
Table 3.4-2 Line Number (236)
Table 3.4-2 Line Number (238)
Table 3.4-2 Line Number (239)
Table 3.4-2 Line Number (240)
Table 3.4-2 Line Number (241)
Table 3.4-2 Line Number (242)
Table 3.4-2 Line Number (243)
Table 3.4-2 Line Number (245)
Table 3.4-2 Line Number (246)
Table 3.4-2 Line Number (247)
Table 3.4-2 Line Number (248)
Table 3.4-2 Line Number (249)
Table 3.4-2 Line Number (250)
Table 3.4-2 Line Number (252)
PULSATION DAMPER PRESSURE BOUNDARY Table 3.4-2 Line Number (253)
Table 3.4-2 Line Number (254)
Table 3.4-1 Line Number (4)
PUMP CASING PRESSURE BOUNDARY Table 3.4-2 Line Number (2) 1 1 1Table 3.4-2 Line Number (270)
Page 2-97
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-1 Chemical and Volume Control (CVCS)
Component Group Passive Function Aging Management Reference Table 3.4-1 Line Number (5)
Table 3.4-2 Line Number (336)
Table 3.4-2 Line Number (338)
Table 3.4-2 Line Number (343)
Table 3.4-2 Line Number (344)
TANK PRESSURE BOUNDARY Table 3.4-2 Line Number (346)
Table 3.4-2 Line Number (347)
Table 3.4-2 Line Number (348)
Table 3.4-2 Line Number (349)
Table 3.4-2 Line Number (350)
Table 3.4-2 Line Number (351)
Table 3.4-2 Line Number (353)
Table 3.4-2 Line Number (356)
Table 3.4-2 Line Number (357)
Table 3.4-2 Line Number (358)
Table 3.4-2 Line Number (359)
TEMPERATURE Table 3.4-2 Line Number (360)
ELEMENTU PRESSURE BOUNDARY Table 3.4-2 Line Number (361)
Table 3.4-2 Line Number (365)
Table 3.4-2 Line Number (366)
Table 3.4-2 Line Number (367)
Table 3.4-2 Line Number (368)
Table 3.4-2 Line Number (369)
Table 3.4-2 Line Number (370)
Table 3.4-2 Line Number (372)
Table 3.4-2 Line Number (373)
Table 3.4-2 Line Number (374)
Table 3.4-2 Line Number (375)
TRANSMITTER' PRESSURE BOUNDARY Table 3.4-2 Line Number (376)
Table 3.4-2 Line Number (377)
Table 3.4-2 Line Number (378)
Table 3.4-2 Line Number (379)
Table 3.4-2 Line Number (381)
Table 3.4-2 Line Number (382)
Page 2-98
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-1 Chemical and Volume Control (CVCS)
Component Group Passive Function Aging Management Reference Table 3.4-1 Line Number (5)
Table 3.4-2 Line Number (384)
Table 3.4-2 Line Number (385)
Table 3.4-2 Line Number (386)
Table 3.4-2 Line Number (393)
Table 3.4-2 Line Number (394)
Table 3.4-2 Line Number (398)
Table 3.4-2 Line Number (399)
Table 3.4-2 Line Number (400)
Table 3.4-2 Line Number (401)
Table 3.4-2 Line Number (402)
Table 3.4-2 Line Number (403)
Table 3.4-2 Line Number (405)
Table 3.4-2 Line Number (406)
Table 3.4-2 Line Number (407)
Table 3.4-2 Line Number (408)
Table 3.4-2 Line Number (409)
Table 3.4-2 Line Number (410)
Table 3.4-2 Line Number (411)
Table 3.4-2 Line Number (412)
VALVE BODY PRESSURE BOUNDARY Table 3.4-2 Line Number (425)
Table 3.4-2 Line Number (429)
Table 3.4-2 Line Number (432)
Table 3.4-2 Line Number (444)
Table 3.4-2 Line Number (446)
Table 3.4-2 Line Number (447)
Table 3.4-2 Line Number (452)
Table 3.4-2 Line Number (453)
Table 3.4-2 Line Number (454)
Table 3.4-2 Line Number (455)
Table 3.4-2 Line Number (456)
Table 3.4-2 Line Number (457)
Table 3.4-2 Line Number (459)
Table 3.4-2 Line Number (461)
Table 3.4-2 Line Number (462)
Table 3.4-2 Line Number (463)
Table 3.4-2 Line Number (464)
Table 3.4-2 Line Number (465)
Table 3.4-2 Line Number (466)
Table 3.4-2 Line Number (467)
- 1. Selected instruments were conservatively included within the scope of License Renewal. Consideration was given to the consequences of an instrument housing pressure boundary failure. Where an instrument was unisol able from a pressure source and is of sufficient size that a system function would be degraded should the pressure boundary fail, that instrument is included for License Renewal review.
Page 2-99
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 2.3.3.2 Component Cooling Water (CCW)
System Description
The Component Cooling Water (CCW) systems' function is to remove heat from safety related heat exchangers during plant operation, plant cooldown, and postaccident conditions. Components within the CCW system are also credited for use in safe shutdown following some fire events.
The principal components of the Component Cooling Water system are pumps, heat exchanges, the surge tank and the essential piping and valves. A single CCW pump circulates chromated water through parallel flow paths into various components where it picks up heat from other systems and transfers the heat to the service water (SW) system via the component cooling water (CCW) heat exchangers. The surge tank accommodates expansion, contraction, and inleakage of water, and ensures a continuous component cooling water (CCW) supply until a leaking cooling line can be isolated. The component cooling loop serves as an intermediate system between the radioactive fluid systems and the service water (SW) system. Since the component cooling water (CCW) system loop is used as an engineered safety feature, containment isolation valves are not automatically closed. That portion of the loop located outside the containment is not required to be a closed system.
The following fluid systems interface with the Component Cooling Water system:
Reactor Coolant Safety Injection Containment Spray Plant Sampling Waste Disposal Service Water System Function Listing In addition to the System Functions listed above, the Component Cooling Water System also supports additional functions (associated design functions) that may or may not be License Renewal Intended Functions. A comprehensive listing of all functions associated with the system, or specific components contained in the system, is provided in the summary below.
Code J Cri 1 Cri 2 Cri 3 PROVIDE HEAT REMOVAL FROM SAFETY RELATED FPJEQ PTSIATISB HEATEXCHANGERS X I I I I Comment: Components within the Component Cooling Water system perform this primary design system function.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Comment: Components within the Component Cooling Water system associated design system function.
Code Q Cri 1 Cri 2 Cri 3 PROVIDE ELECTRICAL POWER TO SAFETY CLASS FP EQ PTS AT SB 1,2,3 COMPONENTS x Comment: Components within the Component Cooling Water system perform this associated design system function.
Code S Cri 1 Cri 2 Cri 3 SPECIAL CAPABILITY CLASS FUNCTIONS FP EQ PTS AT SB Comment: Components within the Component Cooling Water system perform this associated design system function (augmented quality). For the purposes of License Renewal, components within the Component Cooling Water system that perform special capability class functions are tracked under the Criterion 3 codes (Z1 through Z5).
Code T Cri 1 Cri 2 Cri 3 NON-NUCLEAR SAFETY CLASS FUNCTIONS FPI EQIPTS IATI SB Comment: Components within the Component Cooling Water system perform this associated design system function.
Code Zi Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3 - SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT X EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR FIRE PROTECTION (10 CFR 50.48)
Comment: Components within the Component Cooling Water system perform this associated design system function.
UFSAR Reference Additional Component Cooling Water System details are provided in Section 9.2.2 and Table 6.2-15a of the UFSAR.
License Renewal Drawings The license renewal drawings for the Component Cooling Water System are listed below:
Page 2-101
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 33013-1245 33013-1887 33013-1246,1 33013-1888 33013-1246,2 33013-1889 33013-1250,2 33013-1890 33013-1273.1 33013-1892 33013-1891 33013-1899,1 33013-1251,2 Components Subject to an AMR The component groups for this system that require aging management review are indicated in Table 2.3.3-2 along with each Component Group's passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3.
Table 2.3.3-2 Component Cooling Water (CCW)
Component Group Passive Function Aging Management Reference Table 3.4-2 Line Number (26)
Table 3.4-2 Line Number (32)
CS COMPONENTS PRESSURE BOUNDARY Table 3.4-1 Line Number (13)
Table 3.4-1 Line Number (13)
FASTENERS Table 3.4-1 Line Number (23)
FASTENER JOINT INTEGRITY Table 3.4-2 Line Number (79)
(BOLTING) Table 3.4-2 Line Number (80)
Table 3.4-2 Line Number (81)
Table 3.4-2 Line Number (107)
FLOW ELEMENT PRESSURE BOUNDARY Table 3.4-2 Line Number (108)
Table 3.4-2 Line Number (111)
Page 2-102
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-2 Component Cooling Water (CCW)
Component Group Passive Function Aging Management Reference Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (14)
Table 3.4-2 Line Number (120)
Table 3.4-2 Line Number (130)
Table 3.4-2 Line Number (132)
Table 3.4-2 Line Number (133)
Table 3.4-2 Line Number (151)
Table 3.4-2 Line Number (152)
Table 3.4-2 Line Number (153)
PRESSURE BOUNDARY Table 3.4-2 Line Number (154)
HEAT EXCHANGER These apply to the pressure HEAT TRANSFER boundary passive function.
Table 3.4-1 Line Number (16)
Table 3.4-2 Line Number (137)
Table 3.4-2 Line Number (138)
Table 3.4-2 Line Number (140)
Table 3.4-2 Line Number (141)
Table 3.4-2 Line Number (142)
These apply to the heat transfer passive function.
Table 3.4-2 Line Number (175)
INDICATOR 1 PRESSURE BOUNDARY Table 3.4-2 Line Number (176)
Table 3.4-2 Line Number (179)
Table 3.4-2 Line Number (196)
ORIFICE PRESSURE BOUNDARY Table 3.4-2 Line Number (200)
Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (14)
Table 3.4-2 Line Number (221)
PIPE PRESSURE BOUNDARY Table 3.4-2 Line Number (228)
Table 3.4-2 Line Number (234)
Table 3.4-2 Line Number (235)
Table 3.4-2 Line Number (244)
PUMP CASING PRESSURE BOUNDARY Table 3.4-1 Line Number (14)
Tal3.-LieNmr(2)
Table 3.4-2 Line Number (262)
SWITCH' PRESSURE BOUNDARY Table 3.4-2 Line Number (332)
Table 3.4-2 Line Number (333)
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-2 Component Cooling Water (CCW)
Component Group Passive Function Aging Management Reference Table 3.4-2 Line Number (343)
Table 3.4-2 Line Number (352)
TEMPERATURE PRESSURE BOUNDARY Table 3.4-2 Line Number (354)
ELEMENT' Table 3.4-2 Line Number (362)
TRANSMITTER' PRESSURE BOUNDARY Table 3.4-2 Line Number (373)
Table 3.4-2 Line Number (380)
Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (14)
Table 3.4-2 Line Number (386)
Table 3.4-2 Line Number (394)
Table 3.4-2 Line Number (404)
VALVE BODY PRESSURE BOUNDARY Table 3.4-2 Line Number (425)
Table 3.4-2 Line Number (429)
Table 3.4-2 Line Number (436)
Table 3.4-2 Line Number (444)
Table 3.4-2 Line Number (446)
Table 3.4-2 Line Number (447)
- 1. Selected instruments were conservatively included within the scope of License Renewal. Consideration was given to the consequences of an instrument housing pressure boundary failure. Where an instrument was unisol able from a pressure source and is of sufficient size that a system function would be degraded should the pressure boundary fail, that instrument is included for License Renewal review.
2.3.3.3 Spent Fuel Cooling and Fuel Storage
System Description
The spent fuel pool (SFP) cooling system is designed to remove heat from the SFP, which is generated by stored spent fuel. The heat from the SFP is rejected to the Service Water System. The spent fuel pool is a Seismic Category I design, reinforced-concrete structure totally clad with stainless steel.
The SFP provides structural support to the spent fuel racks. The spacing and materials of construction of the spent fuel racks work in conjunction with the spent fuel pool water chemistry to provide reactivity control. The SFP concrete elements are evaluated within the Auxiliary Building structure.
Page 2-104
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information The principal components of the Spent Fuel Cooling and Fuel Storage system include pumps, tanks, heat exchangers and the essential piping and valves.
Hoses are used to connect the skid-mounted equipment into the system. The new and spent fuel storage racks, and the pool and transfer canal liner are included as components within the Spent Fuel Cooling and Fuel Storage system.
The spent fuel cooling system was originally designed as a single train, non-safety system. The system has been modified to add additional cooling flow paths and equipment. The SFP cooling system now consists of three cooling loops. The primary cooling path is loop B. This loop is safety related, seismically qualified, and functions as the preferred system for ensuring adequate cooling in the SFP. The backup loops include permanently installed loop A and a skid-mounted loop. Together these loops act as a 100% back up to the B loop in that they are capable of removing the decay heat from stored spent fuel and a full core off load.
SFP cooling piping is so arranged that failure of any pipeline does not drain the SFP. To protect against the possibility of complete loss of water in the SFP, the upper suction line penetrates the SFP near the top of the pool. The lower suction line penetrates SFP approximately 5 ft.-4 in. above the top of the fuel racks to preclude the possibility of draining the pool and to ensure a minimum water level of 5 ft.-4 in. above the top of the fuel. The SFP cooling water return line, which terminates at the bottom of the SFP, contains an 0.25-in. vent hole near the normal SFP water level so that the pool water cannot be siphoned.
The clarity and purity of the spent fuel pool water is maintained by passing approximately 60 gpm of the loop flow through a filter and demineralizer.
The original spent fuel storage racks provided capacity for the storage of 210 fuel assemblies. In 1976, the NRC approved the replacement of the original racks with higher density flux trap type. This expanded the storage capability from 210 to 595 fuel assemblies. In 1984, the NRC approved the conversion of six flux trap type racks to high-density fixed poison type racks. This further expanded the storage capacity from 595 to 1016 fuel assemblies. At this point, the spent fuel pool was divided into two regions. Region 1 comprised three flux trap type racks to accommodate a full core off-load. Region 2 consisted of six high-density fixed poison (Boraflex) type racks for the storage of 840 fuel assemblies that satisfied minimum burnup criteria and had cooled for a minimum of 60 days.
Page 2-105
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information In 1998, the NRC approved re-racking the spent fuel pool. This re-rack effort will be done in two phases, reconfigured the pool to accommodate a net increase of 353 locations. This is accomplished by retaining the six existing high-density region 2 racks (840 minus 12 for attachment of new racks = 828 locations) and installing new borated stainless steel (BSS) racks with up to 541 additional storage locations for a total of 1369 storage locations after completion of both phases.
After completion of phase 1 of the re-rack, the pool has three types of racks in two regions. Region 1 contains new high-density flux-trap design BSS racks designated as type 3 for fresh and spent fuel. Region 2 contains the existing Boraflex racks designated as type 1 and new high-density BSS racks designated as type 2. With the completion of phase 1, the pool contains 1321 storage locations. Phase 2 has not yet been performed.
As noted above, Boraflex fixed absorber material is provided in the region 2 type 1 racks of the spent fuel pool. The absorber assemblies are welded in place in each storage cell, thus precluding inadvertent mechanical removal. To address concerns with Boraflex degradation as presented in Generic Letter 96-04, RG&E performed tests in February 1998 of the B-1 0 areal density of 24 representative Boraflex panels in region 2 of the spent fuel pool using the Boron Areal Density Gauge for Evaluating Racks (BADGER). During the testing, degradation beyond the four inch gap assumption of the criticality analysis was noted on selected Boraflex panels. This data indicated that some panels had undergone dissolution beyond expected levels and placed the spent fuel pool in an unanalyzed condition. This event and the results of the associated assessment that was performed were reported to the NRC. In addition, the Technical Specifications were changed to ensure that controls are in place to verify at least 2300 ppm of soluble boron is maintained in the spent fuel pool.
Consequently, Boraflex is not relied upon for reactivity control of the stored spent fuel.
New fuel is delivered by truck to the site in approved containers. The assemblies are removed, inspected, and transferred to the new fuel storage racks using the auxiliary building crane. The storage location on the operating level of the auxiliary building facilitates the unloading of trucks and the transfer of the fuel assemblies. The Seismic Category I storage vault contains specially constructed racks which ensure a minimum 20-in. center-to-center spacing of the new fuel assemblies. This spacing ensures a KEFF less than 0.95 for the accidental full water density flooding scenario and less than 0.98 for the Page 2-106
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information accidental low water density (optimum moderation) flooding scenario. The storage area is located above grade, on the Auxiliary Building operating floor, to help prevent this from occurring. The new fuel storage area is configured to store 12 fuel assemblies.
Seismic Il/I and Heavy Loads (NUREG-0612) interfacing issues with respect to the Spent Fuel Pool and Fuel Storage system are addressed in Auxiliary Building and Fuel Handling equipment evaluations.
The following mechanical systems interface with the Spent Fuel Cooling and Fuel Storage system:
Safety Injection System Chemical and Volume Control Service Water Waste Disposal Treated Water System Function Listing In addition to the System Intended Functions listed above, the Spent Fuel Cooling and Fuel Storage System also contains components which support additional functions (associated design system functions) that may or may not be License Renewal Intended Functions. A comprehensive listing of all functions associated with the system, or specific components contained in the system, is provided in the summary below.
Code J Cri 1 Cri 2 Cri 3 PROVIDE HEAT REMOVAL FROM SAFETY RELATED FP EQIPTSlATI SB HEAT EXCHANGERS XII II Comment: Components within the Spent Fuel Cooling and Fuel Storage system perform this associated design system function.
Code K Cril Cri2 Cri3 PROVIDE PRIMARY CONTAINMENT BOUNDARY FPI EQI PTSATi SB Comment: Components within the Spent Fuel Cooling and Fuel Storage system perform this associated design system function.
Code P Cri 1 Cri 2 Cri 3 ENSURE ADEQUATE COOLING INTHE SPENT FUEL FP EQ PTS AT SB POOL X Comment: Components within the Spent Fuel Cooling and Fuel Storage system perform this primary design system function.
Page 2-107
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Code Q Cri 1 Cri 2 Cri 3 PROVIDE ELECTRICAL POWER TO SAFETY CLASS FP EQ PTS AT SB 1,2,3 COMPONENTS X Comment: Components within the Spent Fuel Cooling and Fuel Storage system perform this associated design system function.
Code R Cri 1 Cri 2 Cri 3 STRUCTURALLY SUPPORT OR HOUSE SAFETY FP EQ PTS AT SB CLASS 1,2,3 COMPONENTS x Comment: Components within the Spent Fuel Cooling and Fuel Storage system perform this primary design system function. In addition to pool cooling, the Spent Fuel Cooling system boundary includes the new and spent fuel storage racks and the spent fuel pool liner.
Code S Cri 1 Cri 2 Cri 3 SPECIAL CAPABILITY CLASS FUNCTIONS FP EQ PTS AT SB Comment: Components within the Spent Fuel Cooling and Fuel Storage system perform this associated design system function (augmented quality). For the purposes of License Renewal, components within the Spent Fuel Cooling and Fuel Storage system that perform special capability class functions are tracked under the System Function code Y (Criterion 2).
Components within the Spent Fuel Cooling and Fuel Storage system perform this associated design system function (augmented quality). For the purposes of License Renewal, components within the Spent Fuel Cooling and Fuel Storage system that perform special capability class functions are tracked under the System Function code Y (Criterion 2).
Code T Cri 1 Cri 2 Cri 3 NON-NUCLEAR SAFETY CLASS FUNCTIONS FP EQ PTS AT SB Comment: Components within the Spent Fuel Cooling and Fuel Storage system perform this associated design system function.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Code Y Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 2 - NON SAFETY FP EQ PTS AT SB RELATED SSC'S WHOSE FAILURE COULD X PREVENT SATISFACTORY ACCOMPLISHMENT OF A SAFETY RELATED FUNCTION Comment: Components within the Spent Fuel Cooling and Fuel Storage system perform this associated design system function. Ginna licensing basis includes one safety related Spent Fuel Cooling and Fuel Storage Loop.
Due to the importance of providing backup cooling, equipment which supports the backup cooling function has been conservatively included in the scope of License Renewal and assigned this associated design system function. The Spent Fuel Cooling and Fuel Storage system contains numerous occurrences of safety class changes at locations containing open valves. These locations are where process line instrument connections transition from pipe to tube at instruments which do not provide safety related information. For purposes of License Renewal, these areas are considered in scope and have been assigned this associated system design function.
UFSAR Reference Additional Spent Fuel Cooling and Fuel Storage System details are provided in Section 9.1.1, Section 9.1.2, Section 9.1.3, and Table 6.2-15a of the UFSAR.
License Renewal Drawings The license renewal drawings for the Spent Fuel Cooling and Fuel Storage System are listed below:
33013-1248 33013-1250,2 Components Subject to an AMR The component groups for this system that require aging management review are indicated in Table 2.3.3-3 along with each Component Group's passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3.
Table 2.3.3-3 Spent Fuel Cooling and Fuel Storage Component Group Passive Function Aging Management Reference CS COMPONENTS PRESSURE BOUNDARY Table 3.4-1 Line Number (13)
Table 3.4-2 Line Number (52)
DEMINERALIZER PRESSURE BOUNDARY Table 3.4-2 Line Number (53)
Table 3.4-2 Line Number (54)
Page 2-109
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-3 Spent Fuel Cooling and Fuel Storage Component Group Passive Function Aging Management Reference Table 3.4-2 Line Number (55)
PRESSURE BOUNDARY Table 3.4-2 Line Number (56)
DIAPHRAGM SEAL Table 3.4-2 Line Number (56)
Table 3.4-1 Line Number (13)
FASTNERSTable 3.4-1 Line Number (23)
FASTENERS JOINT INTEGRITY Table 3.4-2 Line Number (79)
(BOLTING) Table 3.4-2 Line Number (80)
Table 3.4-2 Line Number (81)
Table 3.4-2 Line Number (99)
FILTER HOUSING PRESSURE BOUNDARY Table 3.4-2 Line Number (101)
Table 3.4-2 Line Number (102)
Table 3.4-2 Line Number (108)
FLOW ELEMENT PRESSURE BOUNDARY Table 3.4-2 Line Number (109)
Table 3.4-2 Line Number (110)
Table 3.4-1 Line Number (5)
Table 3.4-2 Line Number (119)
Table 3.4-2 Line Number (144)
Table 3.4-2 Line Number (146)
Table 3.4-2 Line Number (148)
Table 3.4-2 Line Number (149)
B PRESSUREANGR DTable 3.4-2 Line Number (150)
HEAT EXCHANGER These apply to the pressure HEAT TRANSFER boundary passive function.
Table 3.4-2 Line Number (145)
Table 3.4-2 Line Number (147)
These apply to the heat transfer passive function.
Table 3.4-2 Line Number (176)
INDICATOR 1 PRESSURE BOUNDARY Table 3.4-2 Line Number (177)
Table 3.4-2 Line Number (178)
Table 3.4-2 Line Number (236)
PIPE PRESSURE BOUNDARY Table 3.4-2 Line Number (238)
Table 3.4-2 Line Number (239)
Table 3.4-2 Line Number (252)
PULSATION DAMPER PRESSURE BOUNDARY Table 3.4-2 Line Number (253)
Table 3.4-2 Line Number (254)
Page 2-110
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-3 Spent Fuel Cooling and Fuel Storage Component Group Passive Function Aging Management Reference Table 3.4-2 Line Number (270)
PUMP CASING PRESSURE BOUNDARY Table 3.4-2 Line Number (274)
Table 3.4-2 Line Number (275)
Table 3.4-2 Line Number (295)
SPECTACLE FLANGE PRESSURE BOUNDARY Table 3.4-2 Line Number (296)
Table 3.4-2 Line Number (297)
Table 3.4-2 Line Number (306)
STRAINER HOUSING PROVIDE FILTRATION Table 3.4-2 Line Number (307)
Table 3.4-2 Line Number (307)
Table 3.4-1 Line Number (9)
PROVIDE RADIATION Table 3.4-1 Line Number (11)
These apply to both passive SHIELDR functions.
STRUCTURE STRUCTURALSUPPORT Table 3.4-1 Line Number (10) This SR EQUIPMENT applies to the structural support SR equipment passive function.
Table 3.4-2 Line Number (343)
TANK PRESSURE BOUNDARY Table 3.4-2 Line Number (346)
Table 3.4-2 Line Number (347)
TEMPERATURE Table 3.4-2 Line Number (354)
ELEMENTU PRESSURE BOUNDARY Table 3.4-2 Line Number (356)
Table 3.4-2 Line Number (357)
Table 3.4-2 Line Number (394)
Table 3.4-2 Line Number (398)
Table 3.4-2 Line Number (399)
Table 3.4-2 Line Number (430)
Table 3.4-2 Line Number (435)
Table 3.4-2 Line Number (447)
Table 3.4-2 Line Number (452)
Table 3.4-2 Line Number (453)
- 1. Selected instruments were conservatively included within the scope of License Renewal. Consideration was given to the consequences of an instrument housing pressure boundary failure. Where an instrument was unisol able from a pressure source and is of sufficient size that a system function would be degraded should the pressure boundary fail, that instrument is included for License Renewal review.
Page 2-111
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 2.3.3.4 Waste Disposal
System Description
The Waste Disposal System provides equipment necessary to collect, process, and prepare for disposal of potentially radioactive liquid, gaseous, and solid wastes produced as a result of reactor operation. Radioactive fluids entering the waste disposal system are collected in sumps and tanks until determination of subsequent treatment methods can be made. The consequences of a Radioactive release from a subsystem or component are evaluated in UFSAR Section 15.7 which concludes that accidental gaseous and liquid radioactive releases from the Waste Disposal system will not pose a safety hazard to the public relative to 10 CFR 100 releases. The Waste Disposal system contains two environmentally qualified sump pumps, which discharge to the waste holdup tank. The waste holdup tank provides a holdup capacity reserved to abate RHR pump seal failure spillage. Other system tanks contain volumes of liquid, which if spilled, could prevent the satisfactory accomplishment of a safety-related function. Additionally components within the system act in concert with structural features to prevent internal floods from propagating.
The principal components of the Waste Disposal system are the demineralizing systems, the waste gas compressors, tanks and the essential piping, pumps and valves. Liquid wastes requiring cleanup before release are collected and processed by a vendor supplied demineralization system. Gaseous waste is re-used as tank cover gas or stored for decay and subsequent release.
The following fluid systems interface with the Waste Disposal System:
Reactor Coolant Residual Heat Removal Instrument Air Containment Spray Spent Fuel Cooling and Fuel Storage Chemical and Volume Control Component Cooling Water Service Water Safety Injection Page 2-112
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information System Function Listing In addition to the System Functions described above, the Waste Disposal System also contains components which support additional functions (associated system design functions) that may or may not be License Renewal Intended Functions. A comprehensive listing of all functions associated with the system, or specific components contained in the system, is provided in the summary below.
Code J Cri 1 Cri 2 Cri 3 PROVIDE HEAT REMOVAL FROM SAFETY RELATED FP EQIPTS ATISB HEAT EXCHANGERS XII II Comment: Components within the Waste Disposal system perform this associated design system function (Component Cooling and Service Water interface with heat exchangers).
Components within the Waste Disposal system perform this associated design system function.
Components within the Waste Disposal system perform this associated design system function (Waste Disposal interfaces with transfer slot drain).
Code Q Cri 1 Cri 2 Cri 3 PROVIDE ELECTRICAL POWER TO SAFETY CLASS FPI EQPTS IATSB 1,2, 3 COM PONE NTS XI I II Comment: Components within the Waste Disposal system perform this associated design system function.
Code S Cri 1 Cri 2 Cri 3 SPECIAL CAPABILITY CLASS FUNCTIONS FPI EQIPTS IATI SB Comment: Components within the Waste Disposal system perform this associated design system function (augmented quality). For the purposes of License Renewal, components within the Waste Disposal system that perform License Renewal Criterion 3 functions are tracked under the Criterion 3 codes (Z1 through Z5).
Page 2-113
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Code T Cri 1 Cri 2 Cri 3 NON-NUCLEAR SAFETY CLASS FUNCTIONS FP EQ PTS AT SB Comment: Components within the Waste Disposal system perform this associated design system function.
Code Y Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 2- NON SAFETY FP EQ PTS AT SB RELATED SSC'S WHOSE FAILURE COULD X PREVENT SATISFACTORY ACCOMPLISHMENT OF A SAFETY RELATED FUNCTION Comment: Components within the Waste Disposal system perform this associated design system function. (Flood protection, backflow of oil through floor drains)
Code Z2 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3- SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT x EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR ENVIRONMENTAL QUALIFICATION (10 CFR 50.49)
Comment: Components within the Waste Disposal system are designated as Environmentally Qualified (Residual Heat Removal pit sump pumps).
UFSAR Reference Additional Waste Disposal System details are provided in Section 11.2, Section 11.3, Section 11.4, Section 3.4.2, Section 9.3.3, and Table 6.2-15a of the UFSAR.
Page 2-114
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information License Renewal Drawings The license renewal drawings for the Waste Disposal System are listed below:
33013-1246,2 33013-1271 33013-1247 33013-1272,1 33013-1258 33013-1273,2 33013-1261 33013-1276 33013-1262,1 33013-1277,2 33013-1262,2 33013-1887 33013-1265.2 33013-1888 33013-1267 33013-1889 33013-1268 33013-1890 33013-1270,1 33013-1892 33013-1272,2 33013-1895 33013-1273,1 33013-1899,1 33013-1274 33013-1900,1 33013-1278,2 33013-2279,2 33013-1891 33013-2280 33013-2287 33013-2289 33013-1259 33013-2742 33013-1270,2 Components Subject to an AMR The component groups for this system that require aging management review are indicated in Table 2.3.3-4 along with each Component Group's passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3.
Table 2.3.3-4 Waste Disposal Component Group Passive Function Aging Management Reference CS COMPONENTS PRESSURE BOUNDARY Table 3.4-1 Line Number (13)
Table 3.4-1 Line Number (13)
Table 3.4-1 Line Number (23)
FASTENERS JOINT INTEGRITY Table 3.4-2 Line Number (79)
(BOLTING) Table 3.4-2 Line Number (80)
Table 3.4-2 Line Number (81)
Table 3.4-1 Line Number (5)
HEAT EXCHANGER PRESSURE BOUNDARY Table 3.4-1 Line Number (14)
Table 3.4-2 Line Number (132)
Page 2-115
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-4 Waste Disposal Component Group Passive Function Aging Management Reference Table 3.4-2 Line Number (197) This applies to the pressure PRESSURE BOUNDARY boundary passive function.
ORIFICE RESTRICTS FLOW Table 3.4-2 Line Number (199) This applies to both passive functions.
Table 3.4-1 Line Number (5)
Table 3.4-2 Line Number (202)
Table 3.4-2 Line Number (229)
PIPE PRESSURE BOUNDARY Table 3.4-2 Line Number (230)
Table 3.4-2 Line Number (231)
Table 3.4-2 Line Number (236)
Table 3.4-2 Line Number (237)
Table 3.4-2 Line Number (256)
Table 3.4-2 Line Number (258)
Table 3.4-2 Line Number (270)
PUMP CASING PRESSURE BOUNDARY Table 3.4-2 Line Number (272)
Table 3.4-2 Line Number (273)
Table 3.4-2 Line Number (274)
Table 3.4-2 Line Number (275)
Table 3.4-2 Line Number (343)
Table 3.4-2 Line Number (345)
Table 3.4-1 Line Number (5)
Table 3.4-2 Line Number (386)
Table 3.4-2 Line Number (391)
Table 3.4-2 Line Number (393)
Table 3.4-2 Line Number (394)
Table 3.4-2 Line Number (397)
Table 3.4-2 Line Number (398)
Table 3.4-2 Line Number (399)
Table 3.4-2 Line Number (428)
Table 3.4-2 Line Number (429)
Table 3.4-2 Line Number (444)
Table 3.4-2 Line Number (446)
Table 3.4-2 Line Number (447)
Table 3.4-2 Line Number (451)
Table 3.4-2 Line Number (452)
Table 3.4-2 Line Number (453)
Page 2-116
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 2.3.3.5 Service Water (SW)
System Description
The Service Water (SW) system takes suction from the ultimate heat sink and supplies the cooling water used to provide heat removal from safety related heat exchangers. SW is also the normal suction supply to the standby auxiliary feedwater system and an alternate supply to the preferred auxiliary feedwater system where it is used to provide emergency heat removal from the reactor coolant system using secondary heat removal capability. The SW system is also credited for use in safe shutdown following some fires. The SW system provides multiple water source flow paths to ensure the availability of the ultimate heat sink. These flow paths include non-safety related equipment whose failure could prevent the satisfactory accomplishment of a safety related function. Portions of the SW distribution system serving safeguards equipment are designed as Seismic Category I. Other portions of the service water (SW) system serving non-safety loads are designated as nonseismic and are capable of being isolated from the Seismic Category I portion.
The principal components of the SW system are four service water pumps, a single loop supply header, essential isolation valves, and other essential piping including the normal and standby discharge header and the intake piping systems that transports water from the lake to the SW pump suction bay.
The SW system consists of a single loop header supplied by two separate, 100% capacity, safety related pump trains. The loop header supplies the cooling water to safety related and non-safety related components and system heat exchangers inside the containment, auxiliary, intermediate, turbine, and diesel generator buildings. The non-safety related and long-term safety functions (e.g., component cooling water heat exchangers) can be isolated from the loop header through use of redundant motor operated isolation valves.
In addition to supplying cooling water to heat exchangers, the system supplies seal water to the circulating water pumps and the vacuum pumps, flushing water to the traveling screens and makeup water to the fire water storage tank via the fire booster pump.
The following fluid systems interface with the Service Water System:
Safety Injection Waste Disposal Residual Heat Removal Plant Air Containment Spray Spent Fuel Cooling and Fuel Storage Chemical and Volume Control Component Cooling Water Page 2-117
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Auxiliary Feedwater Fire Protection Circulating Water Treated Water Plant Sampling System Function Listing In addition to the System Functions described above, the Service Water System also contains components which support additional functions (associated system design functions) that may or may not be License Renewal Intended Functions. A comprehensive listing of all functions associated with the system, or specific components contained in the system, is provided in the summary below.
Code H Cri 1 Cri 2 Cri 3 PROVIDE EMERGENCY HEAT REMOVAL FROM THE FP EQ PTS AT SB REACTOR COOLANT SYSTEM USING SECONDARY X HEAT REMOVAL CAPABILITY I Comment: Components within the Service Water system perform this primary design system function (emergency source of feedwater).
Code J Cri 1 Cri 2 Cri 3 PROVIDE HEAT REMOVAL FROM SAFETY RELATED IFP EQ PTS AT SB HEAT EXCHANGERS X Comment: Components within the Service Water system perform this primary design system function.
Comment: Components within the Service Water system design system function.
Page 2-118
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Code Q Cri 1 Cri 2 Cri 3 PROVIDE ELECTRICAL POWER TO SAFETY CLASS FP EQ PTS AT SB 1,2,3 COMPONENTS x Comment: Components within the Service Water system perform this associated design system function.
Code S Cri 1 Cri 2 Cri 3 SPECIAL CAPABILITY CLASS FUNCTIONS FP EQIPTS AT ISB Comment: Components within the Service Water system perform this associated design system function (augmented quality). For the purposes of License Renewal, components within the Service Water system that perform special capability class functions are tracked under the Criterion 2 code (Y) and Criterion 3 codes (Z1 through Z5). Control Room cooling, Reg Guide 1.97, Backup Spent Fuel Pool Cooling.
Code T Cri 1 Cri 2 Cri 3 NON-NUCLEAR SAFETY CLASS FUNCTIONS FPIEQ PTSIAT SB Comment: Components within the Service Water system perform this associated design system function. Cooling to non-safety related loads, e.g.,
Hydrogen cooler, Bus Duct coolers, etc.
Code Y Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 2- NON SAFETY FP EQ PTS AT SB RELATED SSC'S WHOSE FAILURE COULD X PREVENT SATISFACTORY ACCOMPLISHMENT OF A SAFETY RELATED FUNCTION Comment: Components within the Service Water system perform this associated design system function. Non-safety components within the Service Water system provide an alternate flow paths to ensure the availability of the ultimate heat sink in the event of loss of circulating water. Service Water also supplies cooling to the skid mounted SFP Heat Exchanger via temporary connections.
Code Z1 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3- SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT X EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR FIRE PROTECTION (10 CFR 50.48)
Comment: Components within the Service Water system perform this associated design system function.
Page 2-119
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information UFSAR Reference Additional Service Water System details are provided in Section 3.3.3.3.7, Section 9.2.1, and Table 6.2-15a of the UFSAR.
License Renewal Drawings The license renewal drawings for the Service Water System are listed below:
33013-1237 33013-1885,1 33013-1238 33013-1885,2 33013-1250,1 33013-1892 33013-1250,2 33013-1894,2 33013-1250,3 33013-1895 33013-1893 33013-1900,1 33013-1908,3 33013-1921 33013-1251,1 33013-1925 33013-1251,2 Components Subject to an AMR The component groups for this system that require aging management review are indicated in Table 2.3.3-5 along with each Component Group's passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3.
Table 2.3.3-5 Service Water (SW)
Component Group Passive Function Aging Management Reference CS COMPONENTS PRESSURE BOUNDARY Table 3.4-1 Line Number (13)
EXPANSION JOINT PRESSURE BOUNDARY Table 3.4-2 Line Number (65)
Table 3.4-1 Line Number (13)
FASTENERS Table 3.4-1 Line Number (23)
FASTENES JOINT INTEGRITY Table 3.4-2 Line Number (79)
(BOLTING) Table 3.4-2 Line Number (80)
Table 3.4-2 Line Number (81)
Table 3.4-1 Line Number (16)
FLOW ELEMENT PRESSURE BOUNDARY Table 3.4-2 Line Number (10)
Table 3.4-2 Line Number (108)
Page 2-120
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-5 Service Water (SW)
Component Group Passive Function Aging Management Reference Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (16)
INDICATOR 1 PRESSURE BOUNDARY Table 3.4-2 Line Number (167)
Table 3.4-2 Line Number (169)
Table 3.4-2 Line Number (176)
Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (16)
Table 3.4-2 Line Number (207)
Table 3.4-2 Line Number (208)
Table 3.4-2 Line Number (210)
Table 3.4-2 Line Number (211)
Table 3.4-2 Line Number (212)
Table 3.4-2 Line Number (220)
Table 3.4-2 Line Number (224)
Table 3.4-2 Line Number (226)
Table 3.4-2 Line Number (235)
Table 3.4-2 Line Number (236)
PUMP CASING PRESSURE BOUNDARY Table 3.4-1 Line Number (16)
Table 3.4-1 Line Number (5)
Table 3.4-2 Line Number (303)
Table 3.4-2 Line Number (304)
PRESSURE BOUNDARY These apply to the pressure STRAINER HOUSING PROVIDE FILTRATION boundary passive function.
Table 3.4-1 Line Number (16) This applies to both passive functions.
STRUCTURE HOUSE, PROTECT Table 3.4-2 Line Number (309)
EQUIPMENT Table 3.4-2 Line Number (310)
Table 3.4-1 Line Number (16)
SWITCH1 PRESSURE BOUNDARY Table 3.4-2 Line Number (331)
Table 3.4-2 Line Number (332)
TEMPERATURE Table 3.4-1 Line Number (16)
ELEMENTU PRESSURE BOUNDARY Table 3.4-2 Line Number (353)
Table 3.4-2 Line Number (354)
Page 2-121
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-5 Service Water (SW)
Component Group Passive Function Aging Management Reference Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (16)
Table 3.4-2 Line Number (393)
Table 3.4-2 Line Number (396)
Table 3.4-2 Line Number (417)
VALVE BODY PRESSURE BOUNDARY Table 3.4-2 Line Number (421)
Table 3.4-2 Line Number (422)
Table 3.4-2 Line Number (423)
Table 3.4-2 Line Number (429)
Table 3.4-2 Line Number (446)
Table 3.4-2 Line Number (447)
Table 3.4-2 Line Number (449)
- 1. Selected instruments were conservatively included within the scope of License Renewal. Consideration was given to the consequences of an instrument housing pressure boundary failure. Where an instrument was unisol able from a pressure source and is of sufficient size that a system function would be degraded should the pressure boundary fail, that instrument is included for License Renewal review.
2.3.3.6 Fire Protection (FP)
System Description
The Fire Protection system relies on a strategy that includes combustible materials control, fire detection, fire confinement and fire suppression.
Fixed water spray and sprinkler systems, fixed Halon systems, hose lines, along with portable and wheeled extinguishers, provide fire suppression. A fire header supplies water from Lake Ontario, using a motor driven or diesel driven fire pump, to the suppression systems and hose reel stations. There is also a contingency to use a fire department pumper truck is to provide back-up fire suppression ability.
A city water yard loop supplies water from the Town of Ontario to the plant fire hydrants and to the Screenhouse service water pump area sprinkler system.
The city water yard loop is also used as a backup cooling water supply to the emergency diesel generators and the standby auxiliary feedwater pumps in the event that service water is unavailable.
Page 2-122
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information The fire water system can be used as a backup for the service water system supply to spent fuel pool heat exchanger A, the standby spent fuel pool heat exchanger, motor driven auxiliary feedwater pumps, standby auxiliary feedwater pumps, and the diesel generator lube-oil coolers and jacket water heat exchangers via. temporary hoses. Provisions are also available to use firewater to flush debris off the Circulating Water intake travelling screens when a high-pressure wash is required.
The fire system has a protective signaling system that alarms locally in selected parts of the plant and transmits fire alarm, supervisory, and trouble signals to the control room. Fire barriers are located throughout the plant to separate established fire areas from each other and also to separate certain safety areas from the remainder of the plant. These barriers are designed to stop a fire from propagating from one area to the other. Fire areas have been defined, based upon separation of equipment and essential safe shutdown cables, to ensure that at least one train of safe shutdown equipment is kept free from the effects of a fire. Fire prevention and mitigation considerations have been included in the design of ventilation systems, drain systems, lighting systems, communications systems, electrical and instrument cables, oil collection systems, and civil layout and materials selections. Materials selection and transient combustible material controls help ensure that the Fire Protection system is capable of achieving its suppression function.
The principal components of the Fire Protection system are two segregated fire headers, a diesel and a motor driven pump, the fire water tank, hoses and hose stations, hydrants and the essential piping, spray heads, nozzles and valves.
The system also includes two fixed halon gas suppression systems with their corresponding tanks, pipes, valves and nozzles. Detection, signaling, valve alignment and temperature monitoring systems provide information and automatic controls. Fire barriers, penetration seals, ventilation dampers and doors are addressed in a separate commodity group evaluation as described below.
Fire Barrier Commodity Group The fire hazards analysis submitted to the NRC in February 1977 identified the fire barriers in the plant and the requirements for maintaining their integrity. The Fire Protection Program Report (EPM-FPPR) identifies all the barriers currently necessary as well as their requirements. These barrier requirements were determined by the fire loading calculated for each area subject to a potential fire hazard. As a result of this analysis, several design modifications were implemented at the plant including upgrading of the rating of original barriers and installing new barriers. Additional barriers were also installed to achieve Page 2-123
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information compliance with 10 CFR 50 Appendix R requirements. Fire barriers are located throughout the plant to separate major areas from each other and also to separate certain safety-related areas from the remainder of the plant. These are designed to stop a fire from propagating from one area to another or to minimize the effects of a fire. Penetrations in these barriers are sealed with appropriate materials to match the requirements of the barrier and have been evaluated to demonstrate they provide an acceptable level of fire protection.
A fire barrier is a continuous vertical or horizontal membrane, such as a wall or floor/ceiling assembly, that is designed and constructed with a specified fire resistance rating. Fire barriers limit the spread or damage of fire and may restrict the movement of smoke. Some barriers may have protected openings.
The principal types of fire barriers include:
Fire Rated Assembly - A passive fire protection feature that is used to separate redundant fires safe shutdown capabilities. A fire rated assembly includes fire rated walls, floors, ceilings, equipment hatches, stairwells, doors, dampers, and penetration seals.
Fire Rated Penetration Seal - An opening in a fire barrier for the passage of pipe, cable, etc., which has been sealed so as not to reduce the integrity of the fire barrier.
Internal Conduit Seals:
A. Smoke and Hot Gas Seals - Noncombustible seals installed inside conduit openings to prevent the passage of smoke and hot gasses through fire barriers.
These seals may be located at the fire barrier or at the nearest conduit entry on both sides of the fire barrier. Smoke and hot gas seals are not required to have a fire resistance rating equal to the fire barrier they are installed in.
B. Heat and Fire Seals - Fire rated seals installed inside conduits at or in close proximity to the fire barrier. Heat and fire seals have the same or greater fire resistance rating as the fire barrier they are installed in.
Penetration Seal - Materials, devices, or assemblies installed in communicating spaces across barriers, which provide effective sealing against defined environmental exposure criteria to achieve the same functional requirement as that originally intended by the structural member or the barrier.
Fire Wall - A wall having adequate fire resistance and structural stability under fire conditions to accomplish the purpose of subdividing buildings to restrict the spread of fire.
Fire Break (Fire Stop) - A passive fire protection feature of construction intended to limit flame propagation along vertical or horizontal cable tray runs.
Page 2-124
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Fire Damper - A device, installed in the air distribution system, designed to close automatically upon detection of heat, to interrupt migratory air flow, and to restrict the passage of flame.
Fire Door - The door component of a fire door assembly.
Fire Door Assembly - Any combination of a fire door, frame, hardware, and other accessories, that together provide a specific degree of fire protection to the opening.
Fire Wrap - A passive fire and/or heat resistant covering (Hymec Wrap) used to protect or shield safe shutdown circuits.
Fire Proofing - A passive cementitious coating applied to steel to provide fire resistance.
The structural components of fire barriers are evaluated with the civil structure that contains the penetration (e.g. block walls, concrete floors, etc.). That notwithstanding, the Fire Barrier Commodity Group includes fire doors and door frames (including roll up doors) and fire dampers and damper frames, along with any penetration seals that may make up the transition between those elements and a civil structure. Also included in the commodity group is any additional framing and material used to construct fire resistant enclosures within civil structures (e.g. the turbine seal oil unit enclosure).
Fire barriers are constructed, maintained, and inspected in accordance with specific procedures subject to rigorous quality control measures. Penetration and barrier specific details such as materials configuration, depth of sealing materials, and inspection acceptance criteria are documented and controlled in accordance with special procedures and processes.
The Fire Barrier Commodity Group evaluation boundary includes all of the barrier types described above. Although the generic commodity group review uses representative asset numbers to reflect barrier materials, each fire barrier is labeled in the plant with a unique number. Plant procedures and drawings specifically detail the construction, repair and inspection criteria distinctive to the specific application. Plant procedures and drawing also distinguish which barrier is credited in the licensing basis with respect to fire protection and which barrier is installed for commercial property conservation. The aging management review of these barrier configurations is comprised of an aging effect evaluation of the constituent material of construction for each of the barriers that are in-scope to the License Renewal rule.
The following fluid systems interface with the Fire Protection system:
Auxiliary Feedwater Service Water Plant Air Emergency Power Page 2-125
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information System Function Listing In addition to the System Functions listed above, the Fire Protection System also supports additional functions (associated design functions) that may or may not be License Renewal Intended Functions. A comprehensive listing of all functions associated with the system, or specific components contained in the system, is provided in the summary below.
Comment: Components within the Fire Protection system perform this associated design system function.
Code 0 Cri 1 Cri 2 Cri 3 PROVIDE ELECTRICAL POWER TO SAFETY CLASS FP EQ: PTSATS 1,2,3 COMPONENTS X Comment: Components within the Fire Protection system perform this associated design system function.
Code S Cri 1 Cri 2 Cri 3 SPECIAL CAPABILITY CLASS FUNCTIONS FP EQ PTS AT SB Comment: Components within the Fire Protection system perform this associated design system function (augmented quality). For the purposes of License Renewal, components within the Fire Protection system that perform special capability class functions are tracked under the Criterion 2 code (Y) and Criterion 3 codes (Z1 through Z5).
Code T Cri 1 Cri 2 Cri 3 NON-NUCLEAR SAFETY CLASS FUNCTIONS FP EQ PTS AT SB Comment: Components within the Fire Protection system perform this associated design system function (e.g. fill and pressurization components for the fire water storage tank, system drains, etc.).
Components within the Fire Protection system perform specific safety related functions different from and in addition to the system level functions (e.g. safety related position indication for containment isolation valve 9227).
Page 2-126
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Code Z1 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3 - SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT x EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR FIRE PROTECTION (10 CFR 50.48)
Comment: Components within the Fire Protection system perform this primary design system function.
Code Z5 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3 - SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT x EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR STATION BLACKOUT (10 CFR 50.63)
Comment: Components within the Fire Protection system perform this associated design system function.
UFSAR Reference Additional Fire Protection System details are provided in Section 9.5.1, Section 9.5.1.2, Section 9.5.1.3, Section 7.4.4, and Table 6.2-15a of the UFSAR.
Page 2-127
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information License Renewal Drawings The license renewal drawings for the Fire Protection System are listed below:
33013-1242 33013-2348 33013-1607 33013-2349 33013-1866 33013-2350 33013-1867 33013-2351 33013-1870 33013-2352 33013-1885,2 33013-2353 33013-1893 33013-2354 33013-1989 33013-2355 33013-1990,1 33013-2356 33013-1990,2 33013-2357 33013-1991 33013-2359 33013-1992 33013-1871 33013-1993,1 33013-1892 33013-1993.2 33013-1895 33013-2287 33013-1896 33013-2344 33013-2341 33013-2345 33013-2342 33013-2346 33013-2343 33013-2347 Components Subject to an AMR The component groups for this system that require aging management review are indicated in Table 2.3.3-6 along with each Component Group's passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3.
Table 2.3.3-6 Fire Protection (FP)
Component Group Passive Function Aging Management Reference BELL' PRESSURE BOUNDARY Table 3.4-1 Line Number (20)
Table 3.4-2 Line Number (7)
COMPRESSOR Table 3.4-1 Line Number (5)
CASING (included for PRESSURE BOUNDARY Table 3.4-2 Line Number (10) conservatism)
CONTROLLER 1 PRESSURE BOUNDARY Table 3.4-1 Line Number (20)
Table 3.4-2 Line Number (17)
Page 2-128
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-6 Fire Protection (FP)
Component Group Passive Function Aging Management Reference Table 3.4-1 Line Number (20)
Table 3.4-2 Line Number (18)
COOLER PRESSURE BOUNDARY Table 3.4-2 Line Number (21)
Table 3.4-2 Line Number (22)
Table 3.4-2 Line Number (23)
CS COMPONENTS PRESSURE BOUNDARY Table 3.4-1 Line Number (13)
Table 3.4-2 Line Number (43)
CUTTER ASSEMBLY PRESSURE BOUNDARY Tbe342Ln ubr(4 ENGINE CASING PRESSURE BOUNDARY Table 3.4-1 Line Number (20)
Table 3.4-2 Line Number (58)
Table 3.4-1 Line Number (13)
Table 3.4-1 Line Number (23)
FASTENER JOINT INTEGRITY Table 3.4-2 Line Number (79)
(BOLTING) Table 3.4-2 Line Number (80)
Table 3.4-2 Line Number (81)
Table 3.4-2 Line Number (83)
Table 3.4-2 Line Number (85)
FILTER HOUSING PRESSURE BOUNDARY Table 3.4-2 Line Number (86)
Table 3.4-2 Line Number (98)
Table 3.4-2 Line Number (100)
Table 3.4-2 Line Number (103)
FLAME ARRESTOR FLAME SUPPRESSION Table 3.4-2 Line Number (104)
Table 3.4-2 Line Number (104)
FLOW ELEMENT PRESSURE BOUNDARY Table 3.4-1 Line Number (2)
Table 3.4-1 Line Number (20)
Table 3.4-1 Line Number (20) This PRESSURE BOUNDARY applies to both passive functions.
FLOW NOZZLES Table 3.4-2 Line Number (114) This applies to the pressure boundary passive function.
Table 3.4-1 Line Number (5)
GAS CYLINDER PRESSURE BOUNDARY Table 3.4-2 Line Number (5)
Table 3.4-2 Line Number (1115)
Page 2-129
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-6 Fire Protection (FP)
Component Group Passive Function Aging Management Reference Table 3.4-1 Line Number (20)
HAND CONTROL PRESSURE BOUNDARY Table 3.4-2 Line Number (116)
STATION Table 3.4-2 Line Number (117)
Table 3.4-2 Line Number (118)
Table 3.4-1 Line Number (20) This applies to both passive functions.
Table 3.4-2 Line Number HEATREXCHANGER (126) This applies to the pressure HEAT TRANSFER boundary passive function.
Table 3.4-2 Line Number (136) This applies to the heat transfer passive function.
HEATING ELEMENT PRESSURE BOUNDARY Table 3.4-2 Line Number (159)
Table 3.4-2 Line Number (161)
Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (20)
Table 3.4-2 Line Number (172)
INDICATOR 1 PRESSURE BOUNDARY Table 3.4-2 Line Number (173)
Table 3.4-2 Line Number (174)
Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (6)
LEVEL GLASS PRESSURE BOUNDARY Table 3.4-2 Line Number (187)
Tal34-LieNmr(18 Table 3.4-2 Line Number (188)
Table 3.4-2 Line Number (189)
Table 3.4-2 Line Number (190)
Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (20)
ORIFICE PRESSURE BOUNDARY Table 3.4-2 Line Number (195)
Table 3.4-2 Line Number (196)
Table 3.4-2 Line Number (198)
Page 2-130
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-6 Fire Protection (FP)
Component Group Passive Function Aging Management Reference Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (6)
Table 3.4-1 Line Number (20)
Table 3.4-2 Line Number (202)
Table 3.4-2 Line Number (203)
Table 3.4-2 Line Number (207)
Table 3.4-2 Line Number (209)
PIPE PRESSURE BOUNDARY Table 3.4-2 Line Number (212)
Table 3.4-2 Line Number (213)
Table 3.4-2 Line Number (214)
Table 3.4-2 Line Number (215)
Table 3.4-2 Line Number (219)
Table 3.4-2 Line Number (222)
Table 3.4-2 Line Number (224)
Table 3.4-2 Line Number (227)
Table 3.4-1 Line Number (20)
PROTOMATIC PRESSURE BOUNDARY Table 3.4-2 Line Number (25)
Table 3.4-2 Line Number (251)
Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (6)
Table 3.4-1 Line Number (20)
PUMP CASING PRESSURE BOUNDARY Table 3.4-2 Line Number (262)
Table 3.4-2 Line Number (263)
Table 3.4-2 Line Number (269)
Table 3.4-2 Line Number (271)
Table 3.4-1 Line Number (5)
Table 3.4-2 Line Number (285)
Table 3.4-2 Line Number (286)
PRESSURE BOUNDARY Table 3.4-2 Line Number (287)
RELEASE ASSEMBLY Table 3.4-2 Line Number (287)
Table 3.4-2 Line Number (288)
Table 3.4-2 Line Number (289)
SCREEN PRESSURE BOUNDARY Table 3.4-1 Line Number (20)
Table 3.4-2 Line Number (290)
Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (20)
(29)
PRESSURE BOUNDARY Table 3.4-2 Line Number SPECTACLE FLANGE Table 3.4-2 Line Number (293)
Table 3.4-2 Line Number (294)
Page 2-131
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-6 Fire Protection (FP)
Component Group Passive Function Aging Management Reference PRESSURE BOUNDARY Table 3.4-1 Line Number (20)
SPRINKLER HEAD Table 3.4-2 Line Number (298)
PROVIDE FLOW PROVIDE FLOW These apply to both passive fntos functions.
Table 3.4-1 Line Number (6)
Table 3.4-1 Line Number (20)
Table 3.4-2 Line Number (300)
STRAINER HOUSING PRESSURE BOUNDARY Table 3.4-2 Line Number (301)
Table 3.4-2 Line Number (302)
Table 3.4-2 Line Number (303)
Table 3.4-2 Line Number (305)
Table 3.4-1 Line Number (19)
Table 3.4-1 Line Number (25)
Table 3.4-2 Line Number (308)
Table 3.4-2 Line Number (311)
Table 3.4-2 Line Number (312)
Table 3.4-2 Line Number (313)
Table 3.4-2 Line Number (314)
Table 3.4-2 Line Number (315)
Table 3.4-2 Line Number (316)
Table 3.4-2 Line Number (317)
Table 3.4-2 Line Number (318)
STRUCTURE FIRE BARRIER Table 3.4-2 Line Number (319)
Table 3.4-2 Line Number (320)
Table 3.4-2 Line Number (321)
Table 3.4-2 Line Number (322)
Table 3.4-2 Line Number (323)
Table 3.4-2 Line Number (324)
Table 3.4-2 Line Number (325)
Table 3.4-2 Line Number (326)
Table 3.4-2 Line Number (327)
Table 3.4-2 Line Number (328)
Table 3.4-2 Line Number (329)
Table 3.4-2 Line Number (330)
Page 2-132
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-6 Fire Protection (FP)
Component Group Passive Function Aging Management Reference Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (6)
Table 3.4-1 Line Number (20)
Table 3.4-1 Line Number (21)
Table 3.4-2 Line Number (334)
Table 3.4-2 Line Number (339)
Table 3.4-2 Line Number (341)
Table 3.4-2 Line Number (342)
Table 3.4-2 Line Number (343)
Table 3.4-2 Line Number (344)
TRANSMITTER 1 PRESSURE BOUNDARY Table 3.4-1 Line Number (20)
Table 3.4-2 Line Number (371)
Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (6)
Table 3.4-1 Line Number (20)
Table 3.4-2 Line Number (394)
Table 3.4-2 Line Number (395)
Table 3.4-2 Line Number (414)
Table 3.4-2 Line Number (415)
Table 3.4-2 Line Number (416)
Table 3.4-2 Line Number (417)
Table 3.4-2 Line Number (420)
VALVE BODY PRESSURE BOUNDARY Table 3.4-2 Line Number (427)
Table 3.4-2 Line Number (428)
Table 3.4-2 Line Number (429)
Table 3.4-2 Line Number (431)
Table 3.4-2 Line Number (433)
Table 3.4-2 Line Number (437)
Table 3.4-2 Line Number (444)
Table 3.4-2 Line Number (446)
Table 3.4-2 Line Number (447)
Table 3.4-2 Line Number (450)
Table 3.4-2 Line Number (458)
- 1. Selected instruments were conservatively included within the scope of License Renewal. Consideration was given to the consequences of an instrument housing pressure boundary failure. Where an instrument was unisol able from a pressure source and is of sufficient size that a system function would be degraded should the pressure boundary fail, that instrument is included for License Renewal review.
Page 2-133
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 2.3.3.7 Heating Steam
System Description
The Heating Steam system supports habitability and equipment reliability by maintaining plant area temperatures within acceptable bounds. In addition to supporting ventilation functions, heating system also provides process steam for the waste disposal system evaporator. The system does not perform any nuclear safety function. (Note that at one time the heating steam penetrated containment. Those blanked off sections of abandoned pipe are evaluated within the Containment Isolation Components system.) The Heating Steam system contains pressurized, high temperature fluid and has pipe routing and equipment locations in close proximity to safety related equipment. Accordingly, some localized pipe segment and equipment are considered as non-safety components whose failure could prevent the satisfactory accomplishment of a safety function.
The Heating Steam system is categorized a moderate energy system.
Consequently the effects of heating steam pipe breaks have been evaluated.
Evaluations were subsequently performed to ensure the plant could achieve and maintain safe shutdown following postulated system failures. As a results of the evaluation, pipe whip and jet impingement protection was provided for the 6-in. heating steam line riser located on the intermediate floor of the auxiliary building to protect safety-related electrical equipment in the vicinity of the riser. Additionally, heating steam lines were removed from the relay room and air handling room in order to maintain a mild environment for the purpose of environmental qualification of electrical equipment in the rooms. The mitigative equipment is evaluated in the appropriate civil/structural assessment.
As a result of these analysis and modifications, the only portion of the Heating Steam system considered as non-safety components whose failure could prevent the accomplishment of a safety function are those portions of the system contained in the Diesel Generator rooms.
The principal components of Heating Steam are the boiler, tanks, pumps, condensate collection tanks, unit heaters and essential piping and valves. The heating steam is provided from the house boiler, located in the screen house or from a connection in the main steam system. The systems provided with house steam include: unit heaters in the screen house, intermediate building, auxiliary building, turbine building, diesel generator rooms, auxiliary building air handling units, containment purge supply unit, boric acid batch tank, gas stripper, and the boron recycle evaporator.
Page 2-134
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information The following fluid systems interface with the Heating Steam System:
Main and Auxiliary Steam System Function Listing In addition to the System Functions listed above, the Heating Steam System also supports additional functions (associated design functions) that may or may not be License Renewal Intended Functions. A comprehensive listing of all functions associated with the system, or specific components contained in the system, is provided in the summary below.
Code T Cri 1 Cri 2 Cri 3 NON-NUCLEAR SAFETY CLASS FUNCTIONS FPFE-Q PTSIAT SB Comment: Components within the Heating Steam system perform this associated system function (e.g. habitability heating).
Code Y Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 2- NON SAFETY FP EQ PTS AT SB RELATED SSC'S WHOSE FAILURE COULD X PREVENT SATISFACTORY ACCOMPLISHMENT OF A SAFETY RELATED FUNCTION Comment: Components within the Heating Steam system perform this associated design system function. Components within the Heating Steam system pass through the Emergency Diesel Generator rooms and therefore a failure of these components could affect the ability of safety related components to perform their intended functions.
UFSAR Reference Additional Heating Steam System details are provided in Section 9.4.10 and Section 3.6 of the UFSAR.
Page 2-135
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information License Renewal Drawings The license renewal drawings for the Heating Steam System are listed below:
33013-1250,1 33013-1896 33013-1266 33013-1913 33013-1893 33013-1914 33013-1908,3 330113-1916,1 33013-1915 33013-1916,2 33013-1874 33013-1917 33013-1890 33013-2274 Components Subject to an AMR The component groups for this system that require aging management review are indicated in Table 2.3.3-7 along with each Component Group's passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3.
Table 2.3.3-7 Heating Steam System Component Group Passive Function Aging Management Reference Table 3.4-2 Line Number (155)
Table 3.4-2 Line Number (156)
PIPE PRESSURE BOUNDARY Table 3.4-1 Line Number (5)
Table 3.4-2 Line Number (206)
Table 3.4-1 Line Number (5)
STRAINER HOUSING PRESSURE BOUNDARY Table 3.4-2 Line Number (2)
Table 3.4-2 Line Number (299)
PRESSURE BOUNDARY Table 3.4-2 Line Number (3)
TRAP HOUSING Table 3.4-2 Line Number (383)
Table 3.4-2 Line Number (430)
Table 3.4-2 Line Number (440)
Page 2-136
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 2.3.3.8 Emergency Power
System Description
The diesel generating Emergency Power system provides electrical power for safety related components when the preferred power supply is not available.
The Emergency Power sources become available automatically following the loss of the preferred power supply within a time consistent with the requirements of the engineered safety features and the shutdown systems under both normal and accident conditions. Components within the Emergency Power system are also credited for use in safe shutdown following some fires.
Emergency Power system reliability is a critical element in ensuring that the station demonstrates compliance with regulations for station blackout.
Included in the Emergency Power system are two safety related station Emergency Diesel Generators (EDGs) and the Technical Support Center (TSC) diesel generator. Each EGD is capable of automatically starting and sequentially accepting the power requirements of one complete set of safeguards equipment. Each EDG provides the necessary power to cool the core and maintain the containment pressure within the design value for a loss of coolant accident (coincident with a loss of offsite power). The diesels start automatically when loss of voltage is sensed on the bus they supply. The EDGs also start automatically upon receipt of a safety injection signal. The EDGs are normally operated from the control room but EDG A is equipped with a control station that allows the unit to be electrically divorced from the control room and operated locally. The TSC diesel generator can be used to supply a battery charger in order to support vital DC for long term recovery following some fire scenarios.
The principal components of the EDGs include two diesel engines. Each engine is equipped with its own; turbo charger, air start sub-system, lube oil and cooling water sub-systems, fuel oil sub-system, ventilation system and essential piping and valves. (Ventilation requirements are evaluated separately within the ventilation systems.) The TSC diesel requires its own similar sub-systems to function but uses batteries rather than air as a starting mode of force.
The following fluid systems interface with the Emergency Power System:
Service Water Page 2-137
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information System Function Listing In addition to the System Functions described above, the Emergency Power System also contains components which support additional functions (associated system design functions) that may or may not be License Renewal Intended Functions. A comprehensive listing of all functions associated with the system, or specific components contained in the system, is provided in the summary below.
Code J Cri 1 Cri 2 Cri 3 PROVIDE HEAT REMOVAL FROM SAFETY RELATED FP EQ PTS AT SB HEAT EXCHANGERS x Comment: Components within the Emergency Power system perform this associated system function.
Code Q Cri 1 Cri 2 Cri 3 PROVIDE ELECTRICAL POWER TO SAFETY CLASS FP EQ PTSIATISB 1,2,3 COMPONENTS x TSI I Comment: Components within the Emergency Power system perform this primary system function.
Code S Cri 1 Cri 2 Cri 3 SPECIAL CAPABILITY CLASS FUNCTIONS FP EQ PTS AT SB Comment: Components within the Emergency Power system perform this associated design system function (augmented quality). For the purposes of License Renewal, components within the Emergency Power system that perform special capability class functions are tracked under the Criterion 3 codes (Z1 through Z5).
Comment: Components within the Emergency Power system per system function.
Code Z1 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3- SSC'S RELIED FP iEQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT X EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR FIRE PROTECTION (10 CFR 50.48)
Comment: Components within the Emergency Power system perform this associated system function.
Page 2-138
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information UFSAR Reference Additional Emergency Power System details are provided in Section 9.5,4, Section 9.5.5, Section 9.5.6, Section 9.5.7, Section 9.5.8, and Section 8.3.1.1.6 of the UFSAR.
License Renewal Drawings The license renewal drawings for the Emergency Power System are listed below:
33013-1239,1 33013-1250,1 33013-1239,2 33013-2288 Components Subject to an AMR The component groups for this system that require aging management review are indicated in Table 2.3.3-8 along with each Component Group's passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3.
Table 2.3.3-8 Emergency Power Component Group Passive Function Aging Management Reference Table 3.4-1 Line Number (5)
ACCUMULATOR PRESSURE BOUNDARY Table 3.4-1 Line Number (7)
Table 3.4-1 Line Number (7)
Table 3.4-1 Line Number (5)
Table 3.4-2 Line Number (18)
COOLER PRESSURE BOUNDARY Table 3.4-2 Line Number (19)
Table 3.4-2 Line Number (20)
Table 3.4-2 Line Number (24)
Table 3.4-2 Line Number (25)
ENGINE CASING PRESSURE BOUNDARY Table 3.4-1 Line Number (5)
Table 3.4-2 Line Number (57)
Table 3.4-1 Line Number (5)
Table 3.4-2 Line Number (61)
Table 3.4-2 Line Number (62)
EXPANSION JOINT PRESSURE BOUNDARY Table 3.4-2 Line Number (63)
Table 3.4-2 Line Number (68)
Table 3.4-2 Line Number (69)
Table 3.4-2 Line Number (70)
Page 2-139
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-8 Emergency Power Component Group Passive Function Aging Management Reference Table 3.4-1 Line Number (5)
FAN CASING PRESSURE BOUNDARY Table 3.4-2 Line Number (7)
Table 3.4-2 Line Number (76)
FASTENERS Table 3.4-1 Line Number (23)
(BOLTJOINT INTEGRITY Table 3.4-2 Line Number (79)
Table 3.4-2 Line Number (80)
Table 3.4-1 Line Number (5)
Table 3.4-2 Line Number (82)
Table 3.4-2 Line Number (83)
Table 3.4-2 Line Number (84)
Table 3.4-2 Line Number (85)
FILTER HOUSING PRESSURE BOUNDARY Table 3.4-2 Line Number (88)
Table 3.4-2 Line Number (89)
Table 3.4-2 Line Number (90)
Table 3.4-2 Line Number (91)
Table 3.4-2 Line Number (97)
Table 3.4-2 Line Number (99)
GOVERNOR PRESSURE BOUNDARY Table 3.4-1 Line Number (5)
Table 3.4-2 Line Number (121)
Table 3.4-2 Line Number (122)
Table 3.4-2 Line Number (123)
Table 3.4-2 Line Number (128)
Table 3.4-2 Line Number (130)
PRESSURE BOUNDARY Table 3.4-2 Line Number (133)
HEAT EXCHANGER These apply to the pressure HEAT TRANSFER boundary passive function.
Table 3.4-2 Line Number (127)
Table 3.4-2 Line Number (129)
Table 3.4-2 Line Number (131)
These apply to the heat transfer passive function.
Table 3.4-2 Line Number (159)
HEATING ELEMENT PRESSURE BOUNDARY Table 3.4-2 Line Number (160)
Table 3.4-2 Line Number (161)
Table 3.4-2 Line Number (167)
PRESSURE BOUNDARY Table 3.4-2 Line Number (168)
INDICATOR' Table 3.4-2 Line Number (170)
Table 3.4-2 Line Number (171)
Page 2-140
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-8 Emergency Power Component Group Passive Function Aging Management Reference Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (7)
Table 3.4-2 Line Number (180)
Table 3.4-2 Line Number (181)
LEVEL GLASS PRESSURE BOUNDARY Table 3.4-2 Line Number (182)
Table 3.4-2 Line Number (183)
Table 3.4-2 Line Number (184)
Table 3.4-2 Line Number (185)
Table 3.4-2 Line Number (186)
LUBRICATOR PRESSURE BOUNDARY Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (5)
MUFFLER PRESSURE BOUNDARY Table 3.4-2 Line Number (193)
Table 3.4-2 Line Number (194)
ORIFICE PRESSURE BOUNDARY Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (14)
Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (7)
Table 3.4-1 Line Number (14)
Table 3.4-2 Line Number (212)
Table 3.4-2 Line Number (217)
PIPE PRESSURE BOUNDARY Table 3.4-2 Line Number (218)
Table 3.4-2 Line Number (219)
Table 3.4-2 Line Number (223)
Table 3.4-2 Line Number (225)
Table 3.4-2 Line Number (228)
Table 3.4-2 Line Number (233)
Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (7)
Table 3.4-1 Line Number (14)
Table 3.4-2 Line Tal3.-LieNmr(2) Number (260)
PUMP CASING PRESSURE BOUNDARY Table 3.4-2 Line Number (262)
Table 3.4-2 Line Number (264)
Table 3.4-2 Line Number (266)
Table 3.4-2 Line Number (267)
STRAINER HOUSING PRESSURE BOUNDARY Table 3.4-1 Line Number (7)
Table 3.4-1 Line Number (7)
Page 2-141
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-8 Emergency Power Component Group Passive Function Aging Management Reference Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (7)
Table 3.4-2 Line Number (335)
Table 3.4-2 Line Number (337)
Table 3.4-2 Line Number (340)
Table 3.4-2 Line Number (341)
Table 3.4-2 Line Number (354)
TEMPERATURE Table 3.4-2 Line Number (355)
ELEMENT 1 PRESSURE BOUNDARY Table 3.4-2 Line Number (363)
Table 3.4-2 Line Number (364)
Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (7)
Table 3.4-2 Line Number (389)
Table 3.4-2 Line Number (417)
Table 3.4-2 Line Number (418)
Table 3.4-2 Line Number (419)
Table 3.4-2 Line Number (424)
Table 3.4-2 Line Number (427)
Table 3.4-2 Line Number (429)
VALVE BODY PRESSURE BOUNDARY Table 3.4-2 Line Number (431)
Table 3.4-2 Line Number (432)
Table 3.4-2 Line Number (438)
Table 3.4-2 Line Number (439)
Table 3.4-2 Line Number (445)
Table 3.4-2 Line Number (447)
Table 3.4-2 Line Number (448)
Table 3.4-2 Line Number (450)
Table 3.4-2 Line Number (459)
Table 3.4-2 Line Number (460)
- 1. Selected instruments were conservatively included within the scope of License Renewal. Consideration was given to the consequences of an instrument housing pressure boundary failure. Where an instrument was unisol able from a pressure source and is of sufficient size that a system function would be degraded should the pressure boundary fail, that instrument is included for License Renewal review.
Page 2-142
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 2.3.3.9 Containment Ventilation
System Description
The Containment Ventilation Systems function to provide emergency heat removal from the containment atmosphere, to remove radioactive material from the containment atmosphere, and to provide containment pressure control.
Portions of the system function to maintain specific containment concrete temperatures below the threshold where long term aging effects are manifested. Thus the containment ventilation system is considered to contain non-safety related equipment whose failure could prevent the satisfactory accomplishment of a safety function (e.g. penetration cooling). The Containment Ventilation system also contains components used for fire detection and components that are Environmentally Qualified.
Included within the scope of the Containment Ventilation System are the following subsystems:
- a. Containment recirculation cooling and filtration system.
- b. Control rod drive mechanism cooling system.
- c. Reactor compartment cooling system.
- d. Refueling water surface and purge system.
- e. Containment auxiliary charcoal filter system.
- f. Containment post-accident charcoal filter system.
- g. Containment shutdown purge system.
- h. Containment mini-purge system.
- i. Penetration cooling system.
The principal components of the Containment Ventilation System include filters, fans, dampers, valves, heat exchangers and the essential ductwork and piping.
Fire dampers contained in the system are evaluated as a separate commodity group.
The containment recirculation fans, control rod drive mechanism fans, and reactor compartment fans are direct-driven units, each with standby units for redundancy. The fans and motors of these units are provided with vibration detecting devices to detect abnormal operating conditions in the early stages of the disturbance. Each of the associated systems is provided with flow switches to verify existence of air flow in the associated duct system. Dampers in the following systems and ducts are provided with air by dual supply air mains:
primary compartment ducts, dome ducts, containment auxiliary charcoal filter systems, butterfly valves which isolate the post-accident charcoal filters, and Page 2-143
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information containment purge supply and exhaust ducts. Two of the four fans and coolers plus one containment spray pump (i.e. one train of each system) are required to provide sufficient capacity to maintain the containment pressure within design limits after a loss-of-coolant accident or steam line break accident. The containment recirculation fan cooler electrical connections and other equipment in the containment necessary for operation of the system are capable of operating under the environmental conditions following a loss-of-coolant accident.
The control rod drive cooling system consists of fans and ductwork to draw air through the control rod drive mechanism shroud and eject it to the main containment volume. The reactor compartment cooling system consists of a plenum, cooling coils, fans, and ductwork arranged to supply cool air to the annulus between the reactor vessel and the primary shield and to the nuclear instrumentation external to the reactor. The refueling water surface and purge system supplies air to the surface of the refueling cavity and exhaust from the area above the refueling manipulator crane to protect the operators during refueling operations. The containment auxiliary charcoal filter system's purpose is to absorb radioactive iodine vapor and radioactive particles that may occur as a result of normal primary system leakage inside the containment. The containment shutdown purge system is independent of the main auxiliary building exhaust system and includes provisions for both supply and exhaust air. The supply system includes an outside air connection to roughing filters, heating coils, fans, duct system, and supply penetration with a butterfly valve outside containment and a blind flange inside containment. The exhaust system includes an exhaust penetration with a butterfly valve and a blind flange identical to those above, a duct system, a filter bank with high efficiency particulate air and charcoal filters, fans, and a building exhaust vent. The shutdown purge supply and exhaust duct blind flanges inside the containment are closed during MODES 1, 2, 3 and 4. The containment mini-purge system is capable of purging containment during MODES 1 and 2 at a relatively low flow rate (approximately 1500 cfm). The exhaust is through a 6-in. line to the auxiliary building charcoal filters. The penetration cooling system is used to cool hot mechanical containment penetrations. The containment penetration cooling system is designed to prevent the bulk concrete temperature surrounding the penetrations from exceeding 150 0 F.
The following fluid systems interface with the Containment Ventilation System:
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information System Function Listing In addition to the System Functions described above, the Containment Ventilation System also contains components which support additional functions (associated system design functions) that may or may not be License Renewal Intended Functions. A comprehensive listing of all functions associated with the system, or specific components contained in the system, is provided in the summary below.
Code J Cri 1 Cri 2 Cri 3 PROVIDE HEAT REMOVAL FROM SAFETY RELATED FP EQ PTS AT SB HEAT EXCHANGERS X Comment: Components within the Containment Ventilation system perform this associated design system function (Service Water heat removal from Containment Recirculating Fan Coolers, etc.).
Code K Cri 1 Cri 2 Cri 3 PROVIDE PRIMARY CONTAINMENT BOUNDARY FPEQ PTSI ATISB Comment: Components within the Containment Ventilation system perform this associated design system function.
Code L Cri 1 Cri 2 Cri 3 PROVIDE EMERGENCY HEAT REMOVAL FROM FP EQ PTS AT SB PRIMARY CONTAINMENT AND PROVIDE X CONTAINMENT PRESSURE CONTROL Comment: Components within the Containment Ventilation system perform this primary design system function.
Code M Cri 1 Cri 2 Cri 3 PROVIDE EMERGENCY REMOVAL OF FP EQ PTS AT SB RADIOACTIVE MATERIAL FROM THE PRIMARY X CONTAINMENT ATMOSPHERE Comment: Components within the Containment Ventilation system perform this primary design system function.
Code 0 Cri 1 Cri 2 Cri 3 MAINTAIN EMERGENCY TEMPERATURES WITHIN FP EQ PTS AT SB AREAS CONTAINING SAFETY CLASS 1,2,3 X COMPONENTS Comment: Components within the Containment Ventilation system perform this primary design system function.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Code 0 Cri 1 Cri 2 Cri 3 PROVIDE ELECTRICAL POWER TO SAFETY CLASS FP EQ IPTS AT SB 1, 2, 3 COMPONENTS X Comment: Components within the Containment Ventilation system perform this associated design system function.
Code S Cri 1 Cri 2 Cri 3 SPECIAL CAPABILITY CLASS FUNCTIONS FPI EQIPTS AT SB Comment: Components within the Containment Ventilation system perform this associated design system function (augmented quality). For the purposes of License Renewal, components within the Containment Ventilation system that perform special capability class functions are tracked under the Criterion 2 code (Y) and Criterion 3 codes (Z1 through Z5). Augmented quality equipment in the ventilation system includes Recirc Fan Condensate Level, Containment Purge, RG-1.97 Category 2 Post Accident monitoring variables, etc.
Code T Cri 1 Cri 2 Cri 3 NON-NUCLEAR SAFETY CLASS FUNCTIONS FPI EQIPTS IATISB Comment: Components within the Containment Ventilation system perform this associated design system function.
Code X Cri 1 Cri 2 Cri 3 SFR FUNCTION NOT APPLICABLE AT COMPONENT FPIEQIPTS IATISB LEVEL XII II Comment: Components within the Containment Ventilation system perform specific safety related functions different from and in addition to the system level functions (e.g. supplementary breakers for fans to electrical protect penetrations).
Code Y Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 2- NON SAFETY FP EQ PTS AT SB RELATED SSCS WHOSE FAILURE COULD X PREVENT SATISFACTORY ACCOMPLISHMENT OF A SAFETY RELATED FUNCTION Comment: Components within the Containment Ventilation system perform this associated design system function. Containment penetration and reactor compartment cooling fans maintain containment concrete temperature during normal operation below the temperature at which damage occurs.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Code Z1 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3- SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT x EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR FIRE PROTECTION (10 CFR 50.48)
Comment: Components within the Containment Ventilation system perform this associated design system function (temperature elements).
Code Z2 Cri 1 Cri 2 Cri 3 LICENSE RENEWAL CRITERION 3- SSC'S RELIED FP EQ PTS AT SB UPON IN SAFETY ANALYSES OR PLANT X EVALUATIONS TO PERFORM A FUNCTION THAT DEMONSTRATES COMPLIANCE WITH THE COMMISSION'S REGULATIONS FOR ENVIRONMENTAL QUALIFICATION (10 CFR 50.49)
Comment: Components within the Containment Ventilation system are designated as Environmentally Qualified (Fan cooler motors, containment temperature, etc.).
UFSAR Reference Additional Containment Ventilation System details are provided in Section 6.2.2, Section 6.5.1.2, Section 9.4.1, and Table 6.2-15a of the UFSAR.
License Renewal Drawings The license renewal drawings for the Containment Ventilation System are listed below:
33013-1250,1 33013-1870 33013-1250,3 33013-1884,1 33013-1261 33013-1884,2 33013-1278,1 33013-1893 33013-1863 33013-1915 33013-1864 33013-1887 33013-1865 33013-1888 33013-1866 33013-1916,1 Components Subject to an AMR The component groups for this system that require aging management review are indicated in Table 2.3.3-9 along with each Component Group's passive function(s) and reference(s) to the corresponding AMR Table(s) in Section 3.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-9 Containment Ventilation Component Group Passive Function Aging Management Reference Table 3.4-1 Line Number (5)
Table 3.4-2 Line Number (1)
AIR OPERATED PRESSURE BOUNDARY Table 3.4-2 Line Number (2)
DAMPER HOUSING Table 3.4-2 Line Number (3)
Table 3.4-2 Line Number (4)
Table 3.4-2 Line Number (6)
Table 3.4-2 Line Number (34)
Table 3.4-2 Line Number (35)
Table 3.4-2 Line Number (36)
PRESSURE BOUNDARY Table 3.4-2 Line Number (37)
COOLING COIL Table 3.4-2 Line Number (37)
Table 3.4-2 Line Number (40)
Table 3.4-2 Line Number (41)
CS COMPONENTS PRESSURE BOUNDARY Table 3.4-1 Line Number (13)
Table 3.4-1 Line Number (5)
DAMPER Table 3.4-2 Line Number (47)
HOUSING/FRAME Table 3.4-2 Line Number (48)
Table 3.4-2 Line Number (50)
Table 3.4-1 Line Number (5)
Table 3.4-2 Line Number (59)
EXPANSION JOINT PRESSURE BOUNDARY Table 3.4-2 Line Number (60)
Table 3.4-2 Line Number (66)
Table 3.4-2 Line Number (67)
FAN CASING PRESSURE BOUNDARY Table 3.4-1 Line Number (5)
Table 3.4-1 Line Number (13)
Table 3.4-1 Line Number (23)
FASTENERS JOINT INTEGRITY Table 3.4-2 Line Number (79)
(BOLTING) Table 3.4-2 Line Number (80)
Table 3.4-2 Line Number (81)
Table 3.4-1 Line Number (5)
Table 3.4-2 Line Number (92)
FILTER HOUSING PRESSURE BOUNDARY Table 3.4-2 Line Number (93)
Table 3.4-2 Line Number (97)
Table 3.4-2 Line Number (98)
Table 3.4-1 Line Number (5)
F Table 3.4-2 Line Number (105)
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-9 Containment Ventilation Component Group Passive Function Aging Management Reference Table 3.4-1 Line Number (5) This applies to both passive functions.
Table 3.4-2 Line Number (124)
Table 3.4-2 Line Number (125)
Table 3.4-2 Line Number (143)
PRESSURE BOUNDARY Table 3.4-2 Line Number (146)
HEAT EXCHANGER These apply to the pressure HEAT TRANSFER boundary passive function.
Table 3.4-2 Line Number (134)
Table 3.4-2 Line Number (135)
Table 3.4-2 Line Number (139)
These apply to the heat transfer passive function.
Table 3.4-1 Line Number (5)
HVAC EQUIPMENT PRESSURE BOUNDARY Table 3.4-2 Line Number (162)
PACKAGE 1 Table 3.4-2 Line Number (163)
Table 3.4-2 Line Number (164)
Table 3.4-1 Line Number (2)
Table 3.4-1 Line Number (5)
PIPE PRESSURE BOUNDARY Table 3.4-2 Line Number (232)
Table 3.4-2 Line Number (235)
Table 3.4-2 Line Number (236)
Table 3.4-1 Line Number (5)
Table 3.4-2 Line Number (386)
Table 3.4-2 Line Number (387)
Table 3.4-2 Line Number (392)
Table 3.4-2 Line Number (394)
Table 3.4-2 Line Number (413)
Table 3.4-2 Line Number (417)
Table 3.4-2 Line Number Tal3.-LieNmr(45 (420)
VALVE BODY PRESSURE BOUNDARY Table 3.4-2 Line Number (426)
Table 3.4-2 Line Number (428)
Table 3.4-2 Line Number (429)
Table 3.4-2 Line Number (444)
Table 3.4-2 Line Number (445)
Table 3.4-2 Line Number (446)
Table 3.4-2 Line Number (447)
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 2.3.3-9 Containment Ventilation Component Group Passive Function Aging Management Reference VENTILATION PRESSURE BOUNDARY Table 3.4-2 Line Number (468)
DUCTWORK Table 3.4-2 Line Number (469)
- 1. HVAC equipment packages include the pressure boundary attributes associated with the package and sub-components such as filter housings, internal damper housings, and fan housings. Both the HVAC package units and their associated sub-components are uniquely identified on plant drawings.
2.3.3.10 Essential Ventilation Systems
System Description
The Essential Ventilation Systems functions to maintain temperatures within specified limits in areas containing safety related equipment. Additionally, the control room emergency air-treatment portion of the system is designed to filter the control room atmosphere during periods when the control room is isolated and to maintain radiation levels in the control room at acceptable levels following design-basis accidents. (Note: Radiation detection and toxic gas monitoring are evaluated within the radiation monitoring system.) Ventilation is also required for emergency diesel generator operation, for the technical support center diesel generator and its associated equipment, and the standby auxiliary feedwater pumps, all of which maybe used for safe shutdown following some fire events.
Included within the scope of the Essential Ventilation Systems are the following subsystems:
- a. Auxiliary Building Ventilation
- b. Intermediate Building Ventilation
- c. Standby Auxiliary Feedwater Building Ventilation
- d. Diesel Generator Building Ventilation
- e. Control Building Ventilation
- f. Technical Support Center Ventilation The principal components of the Essential Ventilation Systems include filters, fans, dampers, valves, heat exchangers, conditioning/chiller packages and the essential ductwork and piping. Fire dampers contained in the system are evaluated as a separate commodity group.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information The auxiliary building has a non-safety heating, ventilation, and air conditioning system which provides clean, filtered, and tempered air to the operating floor of the auxiliary building and to the surface of the decontamination pit and spent fuel storage pool. The system exhausts air from the equipment rooms and open areas of the auxiliary building, and from the decontamination pit and spent fuel pool (SFP) through a closed exhaust system. The exhaust system includes a 100%-capacity bank of high efficiency particulate air filters and redundant 100%-capacity fans discharging to the atmosphere via the plant vent. This arrangement ensures the proper direction of air flow for removal of airborne radioactivity from the auxiliary building. Included in the auxiliary building exhaust system is a separate charcoal filter circuit, which exhausts from rooms where fission-product activity may accumulate during MODES 1 and 2 in concentrations exceeding the average levels expected in the rest of the building. Although no credit for this system is assumed in the plant safety analysis, this circuit is capable of providing exhaust ventilation from the areas containing pumps and related piping and valving which are used to recirculate containment sump liquid following a LOCA. A full-flow charcoal filter bank is provided in the circuit, along with two 50%-capacity exhaust fans. The air-operated suction and discharge dampers associated with each fan are interlocked with the fan such that they are fully open when the fan is operating and fully closed when the fan is stopped. These dampers fail to the open position on loss of control signal or control air. The fans discharge to the main auxiliary building exhaust system containing the high efficiency particulate air (HEPA) filter bank. To ensure a path for the charcoal (and HEPA) filtered exhaust to the plant vent if the main exhaust fans are not operating, a fail-open damper is installed in a bypass circuit around the two main exhaust fans. In addition to the main auxiliary building ventilation system (ABVS), the residual heat removal, safety injection, containment spray, and charging pump motors are provided with additional cooling provisions when the pumps are operating.
The safety injection and containment spray pump motors are located in an open area in the basement of the auxiliary building and share three service-water-cooled heat exchangers. In 1992, service water to these heat exchangers was blanked off. The charging pumps and residual heat removal pumps are located in individual rooms, each room being provided with two cooling units consisting of redundant fans, water-cooled heat exchangers, and ductwork for circulating the cooled air. The capacity of each charging pump cooling unit is sufficient to maintain acceptable room-ambient temperatures with the minimum number of pumps required for system operation in service.
The cooling units in the residual heat removal pump pit are not required for the operation of the residual heat removal pumps, even if both pumps are operating. In the event of a loss of offsite power, the ABVS main supply and Page 2-151
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information exhaust fans would be inoperable. However, all other fans in the ABVS are supplied by emergency diesel power, including the pump cooling circuits for safety-related pump motors, as described above. Analysis has shown that the three levels of the auxiliary building and the residual heat removal pump pit would remain within acceptable limits when the outside air was at its maximum expected temperature and there were no cooling units operating. Since the auxiliary building is a very large volume building, it is not expected that there would be a significant postaccident temperature increase except in some local areas near hot piping and large motors. The spent fuel pool (SFP) area ventilation system is a part of the ABVS. The system serves to control airborne radioactivity in the SFP area during normal operating conditions. This is accomplished by directing air from the auxiliary building supply air unit across both the SFP and the decontamination pit to exhaust air ducts which are connected to the suction of the auxiliary building exhaust fan C. Exhaust air from the SFP water surface is drawn through roughing filters and, depending on system alignment, charcoal filters. Discharge from the auxiliary building exhaust fan C passes through HEPA filters, a main auxiliary building exhaust fan, and then out the plant vent.
The non-safety intermediate building ventilation system includes a supply fan that exhausts air from the intermediate building cleanside to the intermediate building restricted area side. Two additional exhaust fans, which are located in the intermediate building restricted area side, draw ventilation air from various areas of both the clean and restricted area sides of the intermediate building and discharge to the auxiliary building discharge header plant vent duct.
Ventilation air is provided to the intermediate building cleanside through louvered outside air intakes, which are located in the east wall of the intermediate building. Additional ventilation air capability is available to be drawn into the intermediate building cleanside from the turbine building through a louvered wall opening, which is installed in front of a rolling fire door installed in the fire barrier wall.
The standby auxiliary feedwater pump (SAFW) room cooling and heating system provides cooling and heating as required to maintain the pump room temperature within the design temperature range. This cooling and heating system is needed to provide an acceptable environment for the equipment in the pump room, which includes the two SAFW and their electric drive motors.
The SAFW room cooling system is capable of operation whenever the SAFW are needed for operation. This is a result of the fact that the cooling system provides the air-cooling required for continuous operation of the pump motors.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information A given cooling unit is automatically started whenever its corresponding SAFW is started. Due to its safety-related nature, the cooling system must remain functional during all modes of plant operation including the period during and after a safe shutdown earthquake.
The diesel generators are housed in adjacent but separate rooms, each room serviced by a safety-related ventilation system. Each room has two inlet fans supplying outside air. Each fan takes suction from a common header and discharges through separate ductwork, dampers, and discharge diffusers. One fan in each room discharges a supply of air directly on the instrument and control cabinets. Excess air is discharged to the outdoors through automatic, pressure-actuated room vents, backdraft dampers, and wall-mounted louvers.
No refrigeration or service water (SW) air-cooling is used.
The control room ventilation system is normally operated using a large percentage of recirculated air. The fresh air intake can be closed to control the intake of airborne activity if monitors indicate that such action is appropriate.
The control room emergency air-treatment system is designed to filter the control room atmosphere during periods when the control room is isolated and to maintain radiation levels in the control room at acceptable levels following the design-basis accident. This system circulates air from the control room, control room office and kitchen through return air ductwork to a central air conditioning unit located in the air handling room. The air is drawn into the unit through roughing type filters and either heated or cooled as required by electric heating or chilled water coils. Conditioned air is directed back to the rooms through a supply air ductwork system. The entire control room emergency zone air volume is turned over approximately 12 times every hour. During normal operation, fresh makeup air is admitted to this system through an intake louver located in the outside wall of the turbine building, the amount varying between 0% to 25% of the unit flow rate, depending on outside air temperature.
Pneumatically operated dampers can be positioned from the control room to isolate the fresh air intake and to place a separate charcoal filter unit in service.
The charcoal filter unit includes both high efficiency particulate air (HEPA) filters and 2-in. deep charcoal adsorbers for removing radioactive particulates and gaseous iodine from the control room atmosphere. Its capacity is approximately 25% of the system flow rate and the unit is installed in a normally isolated bypass circuit. In the event of high radiation levels in the control room, the control room radiation instrumentation will automatically close the redundant dampers in the fresh air intake duct and the dampers in the return air duct to the turbine building, and will open the damper in the charcoal filter unit inlet duct to allow 2000 cfm of the recirculation air to flow through the HEPA filters and charcoal adsorbers. This signal will also start a separate fan to provide flow Page 2-153
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information through the charcoal filter unit. Until radioactivity in the control room atmosphere is reduced to a safe level, system flow will be in a closed cycle from the control room, with approximately 25% bypass flow through the charcoal filter unit, through the air conditioning unit, and back to the control room. The dampers can also be positioned to permit fresh air makeup to the system through the charcoal filter unit. Since all control room penetrations, including doors, are designed to high leak-tightness standards and the control room is maintained at essentially atmospheric pressure, the infiltration of contaminated air into the control room is limited to a very low rate. The control building ventilation system includes within its boundary battery and relay room ventilation. Supplemental heating and cooling to the battery rooms is provided by a non-seismic air conditioning unit, with associated service water piping, ventilation ductwork, electric heating coil, and fire dampers. The electric heating coil is seismically mounted in the heating, ventilation, and air conditioning unit discharge duct. The unit and associated ductwork and piping are designed to function during all plant modes. Although the overall design is nonseismic, the piping and ductwork are designed to maintain structural integrity during a design-basis earthquake. Each battery room has an ac-powered propeller exhaust fan that takes suction from the area to remove hydrogen gas generated by the batteries. Also, there is a separate emergency dc-powered ventilation system that is manually actuated in the event of low air flow in the ductwork of either of these battery room exhaust fans. The relay room contains two self-contained, water-cooled air-cooling units that maintain a normal room temperature.
The technical support center heating, ventilation, and air conditioning system maintains year-round occupancy comfort levels, the heating, ventilation, and air conditioning system provides personnel protection from airborne radiological contaminants, maintains a positive pressure relative to the outside, and provides cooling, heating, and ventilation required by special areas (e.g.
computer room).
The following fluid systems interface with the Essential Ventilation Systems:
Waste Disposal Plant Air Service Water Heating Steam Chilled Water Page 2-154