ML20058G553
| ML20058G553 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 07/30/1982 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20058G537 | List: |
| References | |
| NUDOCS 8208030360 | |
| Download: ML20058G553 (238) | |
Text
, - . .
y UNITED STATES 0F AMERICA l
NUCLEAR R EkG U, L, A'.T 0 R Y COMMISSION s s.
Ci In the Matter of ) [ \
General Electric Company') Docket No. STri 50-447 Standard Plant ) i j .'.
s-AMENDMENT NO. 5 TO APPLICATI0f1 FDA REVIEW OF 2?8 NUCLEAR ISLAND GENERAL ELECTRIC
~
STANDARD SAFETY ANALYSIS REPORT,(GESSAR II)
General Electric Company, applicant in the above captioned proceeding, hereby files Amendment No. 5 to the,238 Nuclear Island Gener al Electric d Standard Safety Analysis Report (GESSAR II). ,
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Amendment No. 5 further amends GESSAR II by: a
- 1. Furnishing the additional information requested in' tiie Commis!. ion's '
k GESSAR II acceptance review letter dated December 9, 1981' pertaining to environmental qualification of safety related ,
equipment (Questions 270.1, 270.2, 270.3 and 270.4 'of Enclosure 1 and Item 1 of Enclosure 2 to acceptance r'e{iew letter). ,
- 2. Providing answers to 3 additional geotechnicaljquestions contained /
in the acceptance review letter (Questions 271. 2T, 271.25 and 271.26 of EnsluAure 1). (
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- 3. Requiring the appileant to provide Chapter 16 Technic @
Specifications. ,
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- 4. Clarifying portions'of the text where obvious '
discrepancir,s exis,t.
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Respectfully submitted, r , .
,j F-- ;
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O GeneralElectricCompany,,/
r by:
S/G G. Sherwood
/s 's G. G. Sherwood, Magager Nucle,ar' Safety and Licenqng 0peration ,
Subscribed and sworn to befoie me this '30 day of July 19D[. . t .
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, 'a 'f s .5 by: S/Ruthe M J innamon ; 7',
. 4 Ruthe M] Kinnamon Notary Public - California i Santa C}a,ra County, ;' f ,
My Commission Expires 4 +1 .i April 26, 1985 . , .
8208030360 820730 N PDR ADOCK 05000447 175 Curtner Avenue K PDR San Jose, CA 95125 nr JNF:pab/J07231* ,
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GESSAR II 22A7007 f\ 's
,) 238 NUCLEAR ISLAND Rev. 5 s
INSTRUCTIONS FOR FILING AMENDMENT NO. 5 Appendix 31 material will be contained in Volume 8.
Only a single copy of the amended Summary Table of Contents is included in this amendment. Additional copies will be provided upon request.
Remove and insert the pages listed below. Dashes (----) in the remove or insert column indicate no action required.
REMOVE INSERT Summary Table of Contents x, xviii, xix, and xxxii through x, xviii, xix, and xxxii through xxxiv xxxiv i " Chapter 1 1.1-1 and 1.1-2 1.1-1, 1.1-la, and 1.1-2 Appendix IC
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1C.2.1-1/1C.2.1-2 1C . 2.1- 1/1C . 2.1- 2
< Chapter 3 3.ll-i/3.ll-fi through 3.11-i through 3.11-163/3.11-164 3.11-136 3A.3-1, 3A.5-9/3A-5.10, and 3A.3-1, 3A.5-9/3A.5-10, and 3A.8-24 3A.8-24 Appendix 3I
New Appendix 3I Chapter 7 7.1-40, 7.4-41, 7.1-77 7.1-40, 7.1-41, 7.1-77 through through 7.1-80, 7.3-418 7.1-80, 7.3-418, 7.6-104, 7.6-104, 7.6-176, 7.7-165, 7.6-104a, 7.6-176, 7.7-165 and 7.7-166 and 7.7-166 Chapter 16 -
Chapter 16 Revised Chapter 16 7-
\._)
Amendment 5 July 30, 1982 pab/J07234
GESSAR II 2217007 238 NUCLEAR ISLAND Rev. 5 h REMOVE INSERT Chapter 19 19.1.2-1/19.1.2-2, 19.1.3-1, 19.1.2-1/19.1.2-2, 19.1.3-1, 19.1.3-2, 19.1.6-1/19.1.6-2, 19.1.3-2/19.1.6-1/19.1.6-2, 19.1.7-1/19.1.7-2, 19.1,7-1/19.1.7-2 19.1.9-1/19.1.9-2, 19.1.12-1/19.1.12-2, 19.1.9-1/19.1.9-2, 19.3.3.3-1/19.3.3.3-2, 19.3.3.31-1/ 19.1.12-1/19.1.12-2 19.3.3.31-2, 19.3.3.34-1/19.3.3.34-2, 19.3.3.35-1/ 19.3.3.3-1/19.3.3.3-2, 19.3.3.35-2 19.3.3.37-1, 19,3.3.37-2, 19.3.3.31-1 through 19.3.3.38-1/19.3.3.38-2, 19.3.3.31-7/19.3.3.31-8, 19.3.3.39-1/19.3.3.39-2, and 19.3.3.34-1/19.3.3.34-2, 19.3.3.40-1/19.3.3.40-2 19.3.3.35-1/19.3.3.35-2, 19.3.3.37-1/19.3.3.37-2, 19.3.3.38-1/19.3.3.38-2, 19.3.3.39-1/19.3.3.39-2 and 19.3.3.40-1/19.3.3.40-2 O
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l 2- Amendment 5 July 30, 1982 pab/J07234 7/23/82
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
(}
SUMMARY
TABLE OF CONTENTS (Continued) l- Chapter /
Section Title Volume 3.10 SEISMIC QUALIFICATIONS OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT (INCLUDING HYDRODYNAMIC EFFORTS) 5 3.10.1 Seismic Qualification criteria (Including Hydrodynamic Loads) 3.10.2 Methods and Procedures for Qualifying Electrical Equipment and Instrumentation 3.10.3 Methods and Procedure of Seismic Analysis or Testing of Supports of Electrical Equipment and Instrumentation (Including Hydrodynamic Loads) 3.10.4 Seismic Qualification Tests and Analyses (Including Hydrodynamic Loads) 3.11 ENVIRONMENTAL DESIGN OF SAFETY-RELATED MECHANICAL AND ELECTRICAL EQUIPMENT 5
/]
(_/ 3.11.1 -Equipment Identification and Environmental Conditions 3.11.2 Qualification Tests and Analyses 3.11.3 Qualification Results 3.11.4 Loss of Ventilation 3.11.5 Estimated Chemical and Radiation Environment APPENDIX 3A SEISMIC SOIL-STRUCTURE INTERACTION ANALYSIS OF THE NUCLEAR ISLAND 5 APPENDIX 3B CONTAINMENT LOADS 6,7
, APPENDIX 3C COMPUTER PROGRAMS USED IN THE DESIGN OF SEISMIC CATEGORY I STRUCTURES 8 APPENDIX 3D ANALYSIS OF RECIRCULATION MOTOR AND PUMP UNDER ACCIDENT CONDITIONS 8 APPENDIX 3E DESCRIPTION OF SAFETY RELATED MECHANICAL AND ELECTRICAL EQUIPMENT 8 APPENDIX 3F DYNAMIC BUCKLING CRITERIA FOR-CONTAINMENT VESSEL 8 ix
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
SUMMARY
TABLE OF CONTENTS (Continued) g Chapter /
Section Title Volume APPENDIX 3G PIPE FAILURE ANALYSIS 8 APPENDIX 311 EFFECT OF CONCRETE ANNULUS BELOW ELEVATION (-) 5 FT., 3 IN. ON SEISMIC DESIGN LOADS AND BUILDING RESPONSES 8 APPENDIX 3I . ENVIRONMENTAL QUALIFICATION TESTING EFFORT ADMINISTRATIVE CONTROLS 8 __
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
(}
SUMMARY
TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 7 INSTRUMENTATION AND CONTROL SYSTEMS 7.l' INTRODUCTION (All Systems) 12 7.1.1 Identification of Safety-Related Systems 7.1.2 Identification of Safety and Power
- Generation Criteria 7.2 REACTOR PROTECTION (TRIP) SYSTEM (RPS) 12 7.2.1 Description 7.2.2 Conformance Analysis 7.3 ENGINEERED SAFETY FEATURES SYSTEM, t INSTRUMENTATION AND CONTROL 13 7.3.1 Description 7.3.2 Analysis
-HPCS -Shield Building Annu us Mixing 4
-ADS
-Secondary Contain-
-LPCS ment Isolation
-RHR/LPCI -Primary Containment
-CRVICS Isolation LCS
-MSPLCS -Standby Power . _ . ,
-RHR/ Containment -D-G Support Systems Spray -Essential Service
-RHR/ Suppression Pool Water Cooling -ESF Area Cooling i -Suppression Pool -Pneumatic Supply 1
Makeup
-CB Atmospheric
-Combustible Gas Control f Control
-CB Chilled Water
-SGTS 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 14 7.4.1 Description 7.4.2 Analysis
-RCIC -RHR/ Shutdown Cooling
-SLC -Remote Shutdown 1
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
SUMMARY
TABLE OF CONTENTS (Continued)
O Chapter /
Section Title Volume 7.5 SAFETY-RELATED DISPLAY INSTRUMENTATION 14 7.5.1 Description 7.5.2 Analysis
-Nuclenet Control -BOP Benchboard "8 -Supervisory Moni-
-Standby Information toring Console _
Panel
-Rx Core Cooling BB -
7.6 ALL OTHER INSTRUMENTATION SYSTEMS REQUIRED FOR SAFETY 14 7.6.1 Description 7.6.2 Analysis 7.6.3 Additional Design Considerations Analyses 7.6.4 References
-Neutron Monitoring -FPCCS g
-Process Radiation -DW/ Containment Monitoring Vacuum Relief
-Leak Detection -Vent & Pressure ]
Control
-Rod Pattern Control
~^
-HP/LP System Interlock -Suppression Pool
-Re irculation Pump g '#f g
l xviii
. -,.. ._ = .- .- - . _. .- .. .-
- GESSAR II. 22A7007 238 NUCLEAR ISLAND Rav. 5 4
SUMMARY
TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume
.i 7.7 CONTROL SYSTEMS NOT REQUIRED FOR SAFETY 14 7.7.1 Description 7.7.2 Analysis 7.7.3 References
-RPV Instrumentation -Leak Detection .
-Rod Control & -Fire Protection _
Information -Drywell Chiller &
-Recirculation Flow Cooling ontrol -Plant Instrument Air
-Feedwater Control -Neutron Monitoring 1 -Performance Moni- -Display Control System
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toring System
-Refueling Interlocks i -Radwaste -
7.8 NI/ BOP INTERFACES 14 7.8.1 Essential Service Water (Supply) 0 7.8.2 System Instrumentation and Controls Diesel Generator-Fuel Oil Transfer
- System APPENDIX 7A I&C ELEMENTARY DIAGRAMS 15,16,17-)
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4 xix I
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 3
SUMMARY
TABLE OF CONTENTS (Continued)
O Chapter /
Section Title Volume 8 ELECTRIC POWER _
8.1 INTRODUCTION
18 8.1.1 Utility Grid Description 8.1.2 Onsite Electric Power System 8.1.3 Design Bases 8.2 OFFSITE POWER SYSTEM 18 8.2.1 Description 8.2.2 Analysis 8.2.3 Nuclear Island - BOP Interface 8.3 ONSITE POWER SYSTEMS 18 -
8.3.1 AC Power Systems 8.3.2 DC Power Systems 8.3.3 Fire Protection of Cable Systems e
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GESSAR II 22A7007 238' NUCLEAR ISLAND Rev. 4
SUMMARY
TABLE OF CONTENTS (Continued) r O-Chapter /
Section Title Volume t
15.6 DECREASE IN REACTOR COOLANT INVENTORY 23 15.6.1 Inadvertent Safety / Relief Valve Opening 15.6.2 Failure of Small Lines Carrying Primary Coolant Outside Containment 15.6.3 Steam Generator Tube Failure
! 15.6.4 Steam System Piping Break'Outside ,
Containment. l 15.6.5 Loss-of-Coolant Accidents (Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary) - Inside Containment 15.6.6 Feedwater Line Break - Outside 3
Containment 2
15.6.7 References p' s-15.7 RADIOACTIVE RELEASE FROM SUBSYSTEMS AND COMPONENTS 23 15.7.1 Radioactive ..:ste System Leak or.
4 Failure 15.7.2 Liquid Radioactive System Failure ,
, 15.7.3 Postulated Radioactive Releases Due to Liquid Radwaste Tank Failure
! 15.7.4 Fuel-Handling Accident 15.7.5 Spent Fuel Cask Drop Accidents
! 15.7.6 References 15.8 ANTICIPATED TRANSIENTS WITHOUT SCRAM 15.8.1 Requirements 23 15.8.2 Plant Capabilities Additional Modifications 15.8.3 _
APPENDIX 15A PLANT NUCLEAR SAFETY OPERATIONAL ANALYSIS 23 4
APPENDIX 15B BWR/6 GENERIC ROD WITHDRAWAL ERROR ANALYSIS 23 i
APPENDIX 15C FAILURE MODES AND EFFECTS ANALYSES 23A APPENDIX 15D- SEVERE ACCIDENTS 23A t
XXXi-
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
SUMMARY
TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 16 TECIINICAL SPECIFICATIONS 24 16.1 DEFINITIONS O
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O's /
SUMMARY
TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 16.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM 24 SETTINGS
, 16.2.1 Safety Limits
, 16.2.2 Limiting Safety System Settings
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-16.B2 BASES FOR SAFETY LIMITS AND LIMITING SAFETY SETTINGS 24
() 16.B2.1 Safety Limits _
16.B2.2 Limiting Safety System Settings
, 16.3/4 LIMITING CONDITIONS FOR OPERATION AND i SURVEILLANCE REQUIREMENTS 24 16.3/4.0 Applicability 16.3/4.1 Reactivity Control Systems ]'
! 16.3/4.2 Power Distribution Limits f
16.3/4.3 Instrumentation 16.3/4.4 Reactor Coolant System 16.3/4.5 Emergency Core Cooling Systems 16.3/4.6 Containment Systems 16.3/4.7 Plant Systems
- 16.3/4.8 Electrical Power Systems
- 16.3/4.9 Refueling Operations 16.3/4.10 Special Test Exceptions 4 O xxxiii l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
SUMMARY
TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 16.B3/4.0 Applicability 16.B3/4.1 Reactivity Control Systems 16.B3/4.2 Power Distribution Limits 16.B3/4.3 Instrumentation 16.B3/4.4 Reactor Coolant System 16.B3/4.5 Emergency Core Cooling System 16.B3/4.6 Containment Systems 16.B3/4.7 Plant Systems 16.B3/4.8 Electrical Power Systems 16.B3/4.9 Refueling Operations 16.B3/4.10 Special Test Exceptions 16.5 DESIGN FEATURES 24 16.5.1 Site 16.5.2 Containment 16.5.3 Reactor Core 16.5.4 Reactor Coolant System 16.5.5 Meteorological Tower Location 16.5.6 Fuel Storage 16.5.7 Component Cyclic or Transient Limit
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16.6 ADMINISTRATIVE CONTROLS 24 16.6.1 Responsibility 16.6.2 Organization 16.6.3 Unit Staff Qualifications 16.6.4 Training 16.6.5 Review and Audit 16.6.6 Reportable Occurance Action 16.6.7 Safety Limit Violation 16.6.8 Procedures 16.6.9 Reporting Requirements 16.6.10 Record Retention 16.6.11 Radition Protection Program 16.6.12 High Radiation Area _
xxxiv l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O 1. INTRODUCTION AND GENERAL DESCRIPTION OF PLANT
1.1 INTRODUCTION
The General Electric Standard Safety Analysis Report, GESSAR II, is written in accordance with Regulatory Guide 1.70 (Standard Format and Content of. Safety Analysis Reports for Nuclear Power Plants, Revision 3, November 1978). For consistency with NUREG-0800 (Standard Review Plan for the Review of Safety Analy-sis Reports for Nuclear Power Reports, Revision 0, July 1981) ,
GESSAR II includes Section 15.8 which addresses anticipated transients without scram and Chapter 18 which addresses human factors. Finally, GESSAR II contains Chapter 19 to serve as a question and response guide.
The GESSAR II response to TMI-related matters is contained in
() Appendix 1A. The assessment of unresolved safety issues is given in Appendix 1B. _ Appendix lC gives the GESSAR II response to the NRC additional guidance provided in the Commissions' GESSAR II acceptance review letter, dated December 9, 1981. _
l.1.1 Type of License Required This General Electric Standard Safety Analysis Report (GESSAR) is submitted in support of the application for a construction permit and facility operating license for the Nuclear Island portion of a nuclear powered electric generating plant. The Nuclear Island (sometimes referred to as Reactor Island) consists of all buildings which are dedicated exclusively or primarily to housing systems and equipment related to the nuclear system. Under the concept presented herein, there are seven such buildings that comprise the Nuclear Island. These are:
() (1) Reactor Building (including shield building and containment) ;
1.1-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 (2) Fuel Building; (3) Auxiliary Building; (4) Diesel Generator Buildings; (5) Control Building; and (6) Radwaste Building.
The only major system related to the nuclear system that is not housed in one of the seven buildings is the Offgas System which is more appropriately housed in the turbine building since it is physically associated with the condenser air ejectors.
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1.1-la
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 I) 1.1.1 Type of License Required (Continued)
For each of the buildings, the General Electric Company (GE) scope of responsibility includes the design of all structures including the foundation mats and everything within these structures.
System boundaries may vary with the nature of the interface, but the general rule for determining interfaces is that boundaries extend to just outside of the building walls. A major factor in the design process is the determinatirn of the exact description ;
.of the interface. Parameters such as dimensions and orientation of the pipes, type of connections, and pressures at the interface points are established and identified in Section 1.9. It is expected that the Applicant supply will conform to these estab-lished interfaces.
O 1.1.2 Identification of Applicant Applicant will supply.
1.1.3 Number of Plant Units For the purposes of GESSAR II, only a single standard plant will be considered.
1.1.4 Description of Location Applicant will supply.
1.1-2
GESSAR II 22A7007 !
238 NUCLEAR ISLAND Rev. 5
() IC.2 GUIDANCES / RESPONSES 1C.2.1 Environmental Qualification of Safety-Related Equipment GUIDANCE (1)
Commission Memorandum and Order of May 23, 1980 defines the cur-rent staff requirements for qualification of this equipment.
Additional guidance on this matter was provided in a subsequent NRR Order, dated November 26, 1980 (concerning record require-ments), Supplements 2 and 3, dated September 30, 1980 and October 24, 1980, respectively, to IE Bulletin No. 79-OlB, and a generic letter to all holders of cps and OLs, dated October 1, 1980.
RESPONSE (1)
T GESSAR II is committed to NUREG-0588 as implemented by the NRC
] staff. Conformance to NUREG-0588 is discussed in Subsection 3.11.2.1.3. _
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1C.2.1-1/lC.2.1-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
-( ) SECTION 3.11
' CONTENTS Section Title Page 3.11 ENVIRONMENTAL DESIGN OF SAFETY-RELATED MECHANICAL AND ELECTRICAL EQUIPMENT 3.11-1 3.11.1 Equipment Identification and-Environmental Conditions 3.11-2 3.11.1.1 Environmental Conditions for Equipment 3.11-2 3
11.1.2 Equipment Identification 3.11-3 3.11.2 Qualification Tests-and Analyses 3.11-4 3.11.2.1 Qualification Program Basis 3.11-4 3.11.2.1.1 Conformance to 10CFR50, Appendix A 3.11-4 3.11.2.1.1.1 Quality Standards and Records -
Criterion 1 3.11-4 m
3.11.2.1.1.2 Design Bases for Protection Against m Natural Phenomena - Criterion 2 3.11-4 m
- 3.11.2.'l.1.3 Environmental and Missile Design Bases - Criterion 4 3.11-5 3.11.2.1.1.4 Protection System' Failure Modes -
Criterion 23 3.11-5 3.11.2.1.2 Conformance to 10CFR50, Appendix B 3.11-5 3.11.2.1.2.1 Section III - Design Control 3.11-6 3.11.2.1.2.2 Section XI - Test Control 3.11-10 _
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3.11.2.1.3 Conformance to NUREG-0588 3.11-11 3.11.2.1.3.1 Establishment of Qualification Parameters for Design Basis Accidents 3.11-12 3.11.2.1.3.1.1 Temperature and Pressure Conditions Inside Containment Loss-of-Coolant m Accident (LOCA) 3.11-12 m
3.11.2.1.3.1.2 Temperature and Pressure Conditions Inside Containment - Main'Steamline Break (MSLB) 3.11-14 3.11.2.1.3.1.3 Effects of Chemical Spray 3.11-17 3.11.2.1.3.1.4 Radiation Conditions Inside and Outside Containment 3.11-17 _
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- 3.11-i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 CONTENTS (Continued) lh Section Title Page
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3.11.2.1.3.1.5 Environmental Conditions for Outside Primary Containment 3.11-25 3.11.2.1.3.2 Qualification Methods 3.11-26 3.11.2.1.3.2.1 Selection of Methods 3.11-26 3.11.2.1.3.2.2 Qualification by Test 3.11-29 ,
3.11.2.1.3.2.3 Test Sequence 3.11-34 m
3.11.2.1.3.2.4 other Qualification Methods 3.11-34 3.11.2.1.3.3 Margins 3.11-35 3,11.2.1.3.4 Aging 3.11-37 3.11.2.1.3.5 Qualification Documentation 3.11-41 _
3.11.2.2 Class lE Product Environmental -
Qualification Basis 3.11-42 3.11.2.2.1 Scope 3.11-42 3.11.2.2.2 Applicable Documents 3.11-43 3.11.2.2.2.1 General Electric Documents 3.11-43 3.11.2.2.2.2 Codes, Standards, and Regulations 3.11-44 3.11.2.2.3 Reference Documents 3.11-47 3.11.2.2.4 Program Description 3.11-49 3.11.2.2.4.1 General 3.11-49 3.11.2.2.4.2 Type Testing 3.11-50 3.11.2.2.4.2.1 Pretest Inspection 3.11-50 3.11.2.2.4.2.2 Baseline Functional Tests 3.11-51 3.11.2.2.4.2.3 Functional Operating Extremes 3.11-51 3.11.2.2.4.2.4 Product Aging 3.11-51 3.11.2.2.4.2.4.1 General 3.11-51 3.11.2.2.4.2.4.2 Thermal Aging 3.11-52 3.11.2.2.4.2.4.3 Radiation Aging 3.11-54 3.11.2.2.4.2.4.4 Operating Aging 3.11-55 3.11.2.2.4.2.4.5 Vibration Aging 3.11-56 3.11.2.2.f.2.5 Dynamic Event Exposure 3.11-57 3.11.2.2.4.2.5.1 General Requirements for Dynamic Testing 3.11-57 h
3.ll-ii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
('h
%J CONTENTS (Continued)
Section Title Page 3.11.2.2.4.2.5.2 Product and Assembly Testing 3.11-59 3.11.2.2.4.2.5.3 Multiple-Frequency Tests 3.11-61 3.11.2.2.4.2.5.4 Single and Multiaxis Tests 3.11-61 3.11.2.2.4.2.5.5 Time Duration 3.11-62 3.11.2.2.4.2.5.6 Single Frequency Tests 3.11-62 3.11.2.2.4.2.5.7 Fragility Test 3.11-63 3.11.2.2.4.2.5.8 Damping 3.11-63 3.11.2.2.4.2.6 Design Basis Event Exposure 3.11-63 3.11.2.2.4.2.7 Posttest Inspection 3.11-64 3.11.2.2.4.2.8 Modification 3.11-64 3.11.2.2.4.2.9 Repairs 3.11-65 3.11.2.2.4.3 Operating Experience 3.11-65 3.11.2.2.4.3.1 Operating Environment 3.11-65 3.11.2.2.4.3.2 Product Performance 3.11-66
() 3.11.2.2.4.3.3 3.11.2.2.4.4 Qualification Determination Analysis 3.11-66 3.11-67 3.11.2.2.4.4.1 Analytical Techniques 3.11-67 3.11.2.2.4.4.1.1 Dynamic Qualification by Analysis 3.11-68 3.11.2.2.4.4.1.2 General Requirements 3.11-69 3.11.2.2.4.4.1.3 Mathematical Model 3.11-70 3.11.2.2.4.4.1.4 Analysis and Design 3.11-71 3.11.2.2.4.4.1.5 Three Components of Motion 3.11-72 3.11.2.2.4.4.1.6 Static Analysis Method 3.11-73 3.11.2.2.4.4.1.7 Dynamic Analysis Method 3.11-74 3.11.2.2.4.4.1.8 Modal Responses 3.11-74 3.11.2.2.4.4.1.9 Required Response Spectra 3.11-74 3.11.2.2.4.4.1.10 Time History Analysis 3.11-75 3.11.2.2.4.4.1.11 Generation of In-Product Spectra 3.11-75 3.11.2.2.4.4.2 Qualification Determination 3.11-76 3.11.2.2.4.5 Combined Qualification 3.11-76 3.11.2.2.4.6 Ongoing Qualification 3.11-76
() 3.11.2.2.4.7 Margins 3.11-77 ,
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 CONTENTS (Continued) 9 Section Title Page
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3.11.2.2.4.8 Documentation 3.11-78 3.11.2.2.4.8.1 General 3.11-78 3.11.2.2.4.8.2 Environmental / Application Data 3.11-79 3.11.2.2.4.8.3 Functional Requirements 3.11-79 3.11.2.2.4.8.4 Pretest Evaluation 3.11-79 3.11.2.2.4.8.5 Product Analysis Report 3.11-79 3.11.2.2.4.8.6 Product Performance Qualification Specification 3.11.80 3.11.2.2.4.8.7 Test Plan and Procedure 3.11-80 3.11.2.2.4.8.8 Test Report 3.11-80 3.11.2.2.4.8.9 Qualification Report 3.11-81 3.11.3 Qualification Results 3.11-82 3.11.4 Loss of Ventilation 3.11-82 3.11.4.1 Drywell Loss of HVAC 3.11-82 gg 3.11.4.2 Containment Loss of HVAC 3.11-82 3.11.4.3 Auxiliary Building Loss of HVAC 3.11-82 3.11.4.4 Fuel Building Loss of HVAC 3.11-82 3.11.4.5 Turbine Building Loss of HVAC 3.11-83 3.11.4.6 Diesel Generator Building Loss of HVAC 3.11-83 3.11.5 Estimated Chemical and Radiation Environment 3.11-83
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
,O SECTION 3.11 TABLES Table Title Page 3.11-1 Environmental Zones 3.11-85 3.11-2 Environmental Conditions for Reactor Building Equipment 3.11-91 3.11-3 Environmental Conditions and Limits for Auxiliary Building Equipment 3.11-93 3.11-4 Environmental Conditions and Limits for Fuel Building Equipment 3.11-95 3.11-5 Environmental Conditions and Limits for Control Building Equipment 3.11-97 3.11-6 Environmental Conditions and Limits for Diesel Generator Building Equipment 3.11-99 3.11-7 Environmental Conditions and Limits for Turbine Building Equipment 3.11-101 3.11-8 Notes for Tables 3.ll-2 Through 3.11-7 3.11-103 gy 3.11-9 Safety-Related Equipment Identifica-() tion and Environmental Qualification Summary 3.11-105
- 3. ll-9 (B21) Safety-Related Equipment and Environ-mental Qualifications Summary - Main Steam System 3.11-108 3.ll-9(B33) Safety-Related Equipment Identifica-tion and Environmental Qualification Summary - Reactor Recirculation System 3.11-113
- 3. ll-9 (Cll) Safety-Related Equipment Identifica-tion and Environmental Qualification Summary - Rod Control System 3.11-114 3.ll-9(C41) Safety-Related Equipment Identifica-tion and Environmental Qualification Summary - Standby Liquid Control System 3.11.115 3.11-9 (C51) Safety-Related Equipment Identifica-tion and Environmental Qualification Summary - Neutron Monitoring System 3.11-116
- 3. ll-9 (C61) Safety-Related Equipment Identifica-tion and Environmental Qualification Summary - Remote Shutdown System 3.11-117
('; -
3.ll-v
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 TABLES (Continued) lh Table Title Page
- 3. ll-9 (C71) Safety-Related Equipment Identifica- _
tion and Environmental Qualification Summary - Reactor Trip System 3.11-118 3.ll-9(D17) Safety-Related Equipment Identifica-tion and Environmental Qualification Summary - Process Radiation Monitor-ing System 3.11-119
- 3. ll-9 (D2 3 ) Safety-Related Equipment Identifica-tion and Environmental Qualification Summary - Containment Atmosphere Monitoring System 3.11-120
- 3. ll-9 (E12 ) Safety-Related Equipment Identifica-tion and Environmental Qualification Summary - Residual Heat Removal System 3.11-121 3.11-9 (E21) Safety-Related Equipment Identifica-tion and Environmental Qualification Summary - Low Pressure Core Spray System 3.11-123
- 3. ll-9 (E2 2 ) Safety-Related Equipment Identifica-tion and Environmental Qualification Summary - High Pressure Core Spray System 3.11-124
- 3. ll-9 (E31) Safety-Related Equipment Identifica-tion and Environmental Qualification Summary - Leak Detection System 3.11-126 3.ll-9(E32) Safety-Related Equipment Identifica-tion and Environmental Qualification Summary - MS Positive Leakage Control Systems 3.11-131
- 3. ll-9 (E51) Safety-Related Equipment Identifica-tion and Environmental Qualification Summary - Reactor Core Isolation Cooling System 3.11-134
- 3. ll-9 (Fll/ Safety-Related Equipment Identifica-13/15/16/42) tion and Environmental Qualification Summary - Fuel Handling Equipment and Accessories 3.11-136
- 3. ll-9 (G41) Safety-Related Equipment Identifica-tion and Environmental Qualification Summary - Fuel Pool Cooling and Cleanup System 3.11-137 lll 3.ll-vi
GESSAR Il 22A7007 238 NUCLEAR ISLAND Rev. 5
, TABLES (Continued)
Table Title Page i
- 3. ll-9 (H13 ) Safety-Related Equipment Identifica-tion and Environmental Qualification -
Summary - Control Room Panels 3.11-140
- 3. ll-9 (H22) Safety-Related Equipment Identifica-tion and Environmental Qualification Summary - Control Room Panels 3.11-143
- 3. ll-9 (P38) Standby Gas Treatment System 3.11-144
- 3. ll-9 (P4 5) Control Building Chilled Water System 3.11-144
- 3. ll-9 (P53) Pneumatic Supply System 3.11-145
- 3. ll-9 (P60) Water Positive Seal Isolation Leakage Control System 3.11-145
- 3. ll-9 (P61) Air Positive Seal Isolation Leakage Control System 3.11-145
- 3. ll-9 (R4 3) Diesel Emergency Power 3.11-146
- 3. ll-9 (T41) Reactor Building HVAC 3.11-147
- 3. ll-9 (T4 9 ) Safety-Related Equipment Identifica-s j tion and Environmental Qualification Summary - Flammability Control System 3.11-147 3.11-9 (XA3) Diesel Generator Building HVAC 3.11-148
- 3. ll-9 (X63) Fuel Building HVAC 3.11-150
- 3. ll-9 (X73) Auxiliary Building HVAC 3.11-151
- 3. ll-9 (X93) Control Building HVAC 3.11-152 4 3.11-10 Selected GE Positions on NUREG-058 3.11-153 3.11-11 Example of a Test Sequence for Type Testing 3.11-161
, 3.11-12 Recommended Margins 3.11-162 ,
1 4
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O SECTION 3.11 ILLUSTRATIONS Figure Title Page 3.11-1 Drywell and Containment Environmental Zones 3.11-163 O
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GESSAR II 238 NUCLEAR ISLAND 22A7007 Rev. 0 V
3.11 ENVIRONMENTAL DESIGN OF SAFETY-RELATED MECHANICAL AND ELECTRICAL EQUIPMENT This section identifies all the safety-related mechanical and electrical equipment, defines the environmental conditions with respect to limiting design conditions, and documents the qualifica-tion methods and procedures employed to demonstrate the capability to perform safety functions when exposed to those environmental conditions in their respective locations. Seismic qualification is addressed in Sections 3.9 and 3.10 for safety-related mechanical and electrical equipment, respectively.
Limiting design conditions include the following:
(1) Normal Operating Conditions - planned, purposeful, unrestricted reactor operating modes and include
(~T startup, power range, hot standby (condenser available),
\) shutdown, and refueling modes; (2) Abnormal Operating Conditions - any deviation from normal conditions anticipated to occur often enough that design should include a capability to withstand the conditions without operational impairment; (3) Test Conditions - planned testing including pre-operational tests; and (4) Accident Conditions - a single event not reasonably expected during the course of plant operation that has been hypothesized for analysis purposes or postulated from unlikely but possible situations or that has the potential to cause a release of radioactive material (a reactor coolant pressure boundary rupture may qualify T
as an accident; a fuel cladding uefect does not).
N'~J 3.11-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O
3.11.1 Equipment Identification and Environmental Conditions Safety-related mechanical and electrical equipment is located within various environmental zones of the Nuclear Island Buildings and the Turbine Building. Since the environmental conditions the equipment is exposed to within a given building depend on where in the building the equipment is located, there are many potential environmental conditions. This subsection first establishes the environmental conditions for selected zones within each building -
and then identifies the safety-related equipment within them.
3.11.1.1 Environmental Conditions for Equipment There are two groups of Reactor Building equipment environmental zones: the drywell and containment (outside the drywell) (Figure 3.11-1). Each of the other Nuclear Island Buildings housing safety-related mechanical and/or electrical equipment (i.e.,
Auxiliary, Fuel, Control, and Diesel Generator buildings) has one or more environmental zones. All the environmental zones are defined in Table 3.11-1.
With respect to limiting design conditions, Tables 3.11-2 through 3.11-7 define the equipment environmental conditions for each environmental zone and the corresponding duration. The magnitude and 40-year frequency of occurrence of significant deviations from normal plant environments in the zones are listed. Identification of significant enveloping abnormal conditions and each enveloping accident event that impacts the zone environment is included in these tables. Environmental conditions include temperature, pressure, relative humidity, and radiation. The environmental parameters shown do not include margins required to satisfy equip-ment qualification requirements. The environmental conditions shown in these tables are upper-bound envelopes used to establish the environmental design and qualification bases of safety-related equipment. Comparisons of estimated chemical and 3.11-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
,m.
(_) 3.11.1.1 Environment Conditions for Equipment (Continued) radiation environmental conditions with those given in Tables 3.11-2 through 3.11-7 are discussed in Subsection 3.11.5. Table 3.11-8 provides notes referred to in Tables 3.11-2 through 3.11-7.
3.11.1.2 Equipment Identification Table 3.11-9 identifies the safety-related mechanical and electrical equipment. Each table lists the equipment for a particular system by master parts list (MPL) item number and name. The method of qualification, environmental limit (EL) ,
function time (PT), and environmental zone are indicated. The corresponding qualification summary table or qualification report is also identified.
The EL designates the extreme limit for the abnormal and accident
\m ) condition events of Tables 3.11-2 through 3.11-7. The FT corres-ponds to the minimum period of time that safety-related equipment is expected to operate under the extreme limit environment.
The equipment classification code, manufacturer, and model number are tabulated in Appendix 3E.
N __
3.11-3
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GESSAR II. 22AT007 '
238 NUCLEAR ISLAND Rev. S rI
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3.11.2 Oualification Tests and Analyses '
j i .-
3.11.2.1 Qualification Program Basis This subsection describes the conformance of-the GE environmental qualification program to General Design Criteria 1, 2, 4, and 23 of Appendix A of -10CFR50, Sections III and XI of Appendix B of 0 10CFR50, and NUREG-0588. _
3.11.2.1.1 Conformance to 10CFR50, AphendixA _
l.-
3.11.2.1.1.1 Quality Standards and Records - Criterion 1 i Systems and components important to safety are fabricated, erected, and tested to qualify standards commensurate with the safety '#
gg functions to be performed. # s
- /
Section17.1describesthequalityassurance[ program,which ensures that safety-related structures, systems, and components -
satisfactorily perform their safety functions. Section 17.1 also 7 _
describes the documentation which is maintained as evidence of thct' 'd "
' 3 ,
assurance.
- l
+
3.11.2.1.1.2 Design Bases for Protection Ag6inft Natural Phenomena - Criterion 2 _
Structures, systems, and components important to safety are '
designed to withstand the effects of natural phenomena rych as .
earthquakes, tornadoes, hurricanes, floods, tsunami, r;nd seiches without loss of capability to perform thqir safety function. The design bases for these structures, systems, and"comhonents reflect: ,
(1) appropriate consideration of the most severe 3f the natural ',
phenomena that have been historically reported-f r the site and surrounding area, with sufficient margin for the' limited accurac,gi, ,y ll , , I.Y ' , ,
~ . . . .
3.llJ4 ' .
</.
,d o e/
j' <
a
. GESSAR II 22A7007
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230 NUCLEAR ISLAND Rev. 5 O
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l, 3.11.2.1.1.2 Design Bases for Protection Against Natural Phenomena - Criterion 2 (Continued) 2 y
quality, and period of time in which the historical data have been accumulated; (2) appropriate combinations of the effects of b
y normal and accident conditions with the effects of the natural f phenomena; and (3) the safety functions to be performed.
3.11.2.1.1.3 Environmental and Missile Design Bases - Criterion 4 Instrumentation, motors, actuators, and other electrical or j~/ gectronic products designated as Class lE are qualified to meet l j the most adverse accident conditiog to which they would be sub-
, j ] ec'ted'during design basis events. This qualification, preferably ,
.a Phy type testing, subjects the products to the worst case of tem-n
/perature, pressure, radiation, and humidity assumed to exist dur-O, ,
ing the required operational period. As the subject of missiles is outside the scope of NUREG-0588, the GE qualification program does,not address this design basis.
3.11.2.1.1.4 Protection System Failure Modes - Criterion 23 The reactor protection system (RPS) is designed to fail into a safe state, or into a state demonstrated to be acceptable on some f, d-5* #
other defined basis, if conditions such as disconnection of the
/
- .sy. stem, loss of energy (e.g., electric power, instrument aid), or prystulated adverse environments (e.g., extreme heat or cold, pres-J sure,1 steam, water, and radiation) are experienced. j 3.11.2.1.2- Conformance to 10CFR50, Appendix B J'
Qd$lification verification, either by type testing or other I/[ approved means, of Class lE products is conducted under the quality asaurance controls described in Section 17.1 A description of how _
r s ,
[ s*
3.11-5 F /'
.) v _,m 1.,.__ ,y, - - , , , . , , , . . . _ ,
- - . _ , - - , _ . , ,-m. , , , , , _ . _ , .- , , . _ _ . _ . - , , . , . . _ . - ,
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 9
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3.11.2.1.2 Conformance to 10CFR50, Appendix B (Continued) the qualification program complies with Sections III and XI of Appendix B to 10CPR50 is provided in the following paragraphs.
3.11.2.1.2.1 Section III - Design Control The design of structures, systems, and components is controlled within the various design organizations to assure safe and reliable performance of products and services to be supplied. The design control processes are documented in practices and procedures which establish the responsibilities and interfaces of each organiza-tional unit that has an assigned design responsibility. The practices and procedures include measures to assure that:
(1) design requirements are defined and design activities are carried out in a planned, controlled, and orderly manner; (2) appropriate quality requirements and standards are specified in design documents; (3) suitable materials, components, and processes are specified in design documentation; (4) appropriate design verification methods are selected and implemented by individuals or groups having direct responsibility for the original design; and (5) design changes are controlled to the same level as was applied to the original design, including review and approval by the same organization that performed the original review and approval unless another responsible organization is designated by GE management. .
3.11-6 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O '
3.11.2.1.2.1 Section III - Design Control (Continued)
Design specifications and data sheets containing design basis and other data for the Nuclear Island are developed by the design engineer based on the applicable project or program release docu-ment and issued to the responsible design organizations. These design controlling documents provide the basis for detailed system, structure, and component design and typically include the system and structure design specification, piping and instrumentation diagrams, process diagrams, functional control diagrams, and instrument engineering diagrams.
The design specifications, data sheets, and design controlling documents incorporate the design and safety requirements for the Nuclear Island. These designs are subject to independent design verification. The various engineering organizations of GE are responsible for the design and design control activities for the Nuclear Island scope of supply. Engineering personnel are auth-orized to define and prepare performance parameters and to docu-ment the design of systems and equipment. They obtain necessary internal engineering interface consultation and, services as required. They provide final design approval in accordance with documented engineering practices and procedures. Responsibility for internal design document control is vested in the engineering support organizations of GE. Responsibility for interface control with the owner is assigned to the responsible project or program manager.
In addition to the design specifications, data sheets, and design control documents, engineering organizaticus issue general standard i
design specifications which establish standard requirements for designing components which satisfy the system and structures O -
3.11-7
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 9
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3.11.2.1.2.1 Section III - Design Control (Continued) requirements. These standard design specifications identify applicable codes, standards, regulations, and other requirements to be utilized to assure compliance with safety criteria, quality levels, and other specific requirements which have been imposed to obtain acceptable quality, safety, and reliability. Included is the environment and function time to which products must be qualified. These design specifications are subject to design verification review prior to issue.
The design documentation of GE-purchased components (other than instrumentation and controls and materials used in the fabrication of GE-manufactured components) consists of one or more of the following documents: product purchase specifications which specify general requirements, including materials, processes, workmanship, and acceptance criteria; purchased part drawings which show the outline and interface requirements; and specific data sheets or project sheets which define the project-unique requirements of the product. The designs are subjected to design verification review and, in addition, are reviewed by the project manager to assure that the design documents meet any unique project require-ments. Initial issues of purchase specifications for engineered products are reviewed by General Electric's Quality Assurance Engineered Equipment & Installation (QAEE&I) organization prior to supplier bidding.
The purchase specification identifies the documents such as drawings, procedures, calculations, and test data which must be submitted by the supplier for review and approval by Engineering and/or QAEE&I.
Product design and quality requirements are transmitted from Engineering to Purchasing through controlled issuance of material _
3.11-8
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GESSAR II 22A7007 i 238 NUCLEAR ISLAND Rev. 5 O _
3.11.2.1.2.1 Section III - Design Control (Continued) requests (MRs) which identify the product and specify applicable drawings, specifications, and quality assurance (QA) requirements.
Initial issuance and revisions to the MRs are controlled by written procedures.
The system design controlling documentation for controls and instrumentation consists of design specifications, instrument engineering diagrams, functional control drawings, and controls and instrumentation specifications which incorporate the general functional, environmental, material, and test requirements. The responsible engineer obtains interface review of the documents he has initiated and assures that they meet the requirements of the project: applicable nuclear system data sheets, system design control documents, interface control drawings, and general design
() specifications, as well as the applicable codes, standards, and regulations. Reviewers include the project engineer and other engineers responsible for systems or components with which there exists a design interface. In addition, a design verification review is performed to assure correctness and completeness of design, including specification of the appropriate quality requirements.
The detail design of controls and instrumentation by General Electric's engineering and subcontractor designers encompasses the generation of system elementary diagrams and connection dia-grams, panel and rack arrangement drawings, purchased part draw-ings, instrument data sheets, manufacturing drawings, and instruc-l tion manuals. The instrument data sheets define the characteris-tics of the measurable parameters, the instrument environment ranges, accuracies, setpoints, and locations of instruments required by the system design. These detail design documents are ,
3.11-9 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
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0 3.11.2.1.2.1 Section III - Design Control (Continued) subjected to design verification review. The detail design makes use of standard products and purchased components which have been procured in accordance with approved functional specifications and qualified for performance and design adequacy in accordance with documented procedures by a separate testing group. Design requirements, qualification test reports, calculations, and other design data are reviewed for design adequacy with documented pro-cedures. Those documents to be included in the design record file (DRF) are as noted in Subsection 3.11.2.2.4.8.
3.11.2.1.2.2 Section XI - Test Control A product test program is established to provide assurance that all testing required to demonstrate that the product will perform satisfactorily in service is identified, performed, and documented.
Product testing is performed in accordance with QA or engineering test procedures which incorporate or reference the test require-ments and acceptance limits contained in applicable engineering and QA documents. Test requirements and acceptance criteria are normally provided by the organization responsible for the design of the product. The product test program covers all required tests, including, as appropriate, development testing, prototype qualifi-cation testing, calibration of instruments, hydrostatic testing of pressure boundary components, qualification testing of procured components, in-process testing of manufactured components, or final acceptance testing of completed products.
Test procedures for production tests of GE-manufactured products are normally developed by quality control (QC) engineers or cognizant design engineers in accordance with specified test requirements and reviewed by the cognizant design engineer. These O
3.11-10
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 r3 V
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3.11.2.1.2.2 Section XI - Test Control (Continued) test procedures include provisions for assuring that prerequisites for the given test have been met for mandatory hold points (where applicable), for acceptance and rejection criteria, and for appro-priate methods of documenting test results. Prerequisites include such items as appropriate and calibrated test equipment, trained test personnel, completion status of the specimen to be tested, suitable environmental conditions, and provisions for data acqui-sition and storage.
Test procedures for procured products are based on engineering design and ASME code requirements and, when prepared by the sup-plier, are approved by assigned GE engineering or QA personnel.
Product test results are documented, reviewed, and evaluated by 7.-
(-) designated QA engineering personnel prior to release for shipment to assure that test requirements have been satisfied. _
~
3.11.2.1.3 Conformance to NUREG-0588 The following discussion provides an itemized account of the GE interpretation of the NRC staff positions on environmental qual- p ification as published in the 1979 "For Comment" version of NUREG-0588. The format and sequence are consistent with that of NUREG-0588. Each position statement is reprinted as it appears in the staff document, followed by the GE position. Where the GE position differs from that of the staff, justification is provided. ,
o m
3.11-11
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 9
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3.11.2.1.3.1 Establishment of Qualification Parameters for Design Basis Accidents 3.11.2.1.3.1.1 Temperature and Pressure Conditions Iriside Containment Loss-of-Coolant Accident (LOCA) a.(1) NRC Staf f Position 1.1. (l) :
The time-dependent temperature and pressure, established for the design of the containment structure and found acceptable by the staff, may be used for environmental qualification of equipment.
a.(2) GE Position:
The time-dependent temperature and pressure, established for the design of the containment structure and found acceptable lll by the staff, may be used for environmental qualification of p equipment. M b.(l) NRC Staff Position 1.1.(2):
Acceptable methods for calculating and establishing the con-tainment pressure and temperature envelopes to which equipment should be qualified are summarized below. Acceptable methods for calculating mass and energy release rates are summarized in Appendix A.
e Boiling Water Reactors (BWRs) e Mark I, II, and III Containment Calculate LOCA environment using methods of GESSAR Appendix 3B or equivalent industry codes. Additional guidance is provided in SRP Section 6.2.1.1.C, lll NUREG-75/087. ,
3.11-12
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 i
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3.11.2.1.2.1.l' Temperature and Pressure Conditions Inside Contain-ment Loss-of-Coolant Accident (LOCA) (Continued) l b.(2) GE Position:
Guidance for establishing methods for calculating mass'and energy release rates for GESSAR II are provided by the following:
)
tl e Mark III Containment Design NEDO-20533, The General Electric Mark III Pressure Suppression Containment System Analytical Model
, e Other methods found acceptable by the NRC.
a c.(1) NRC Staff Position 1.1.(3):
O In lieu of using the plant-specific containment temperature 4 and pressure design profiles for BWR and ice condenser types of plants, the generic envelope shown in Appendix C may be used for qualification testing.
4 c.(2) GE Position:
The generic envelope shown in Appendix C of NUREG-0588 may be
- used for qualification testing of the product in the drywell in lieu of using plant specific containment temperature and pressure design profiles for BWRs.
d.(1) NRC Staf f Position 1.1. (4) :
The test profiles included in Appendix A to IEEE Std. 323-1974 should not be considered an acceptable alternative in lieu of
() using plant-specific containment temperature and pressure -
3.11-13 1
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
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9 3.11.2.1.2.1.1 Temperature and Pressure Conditions Inside Contain-ment Loss-of-Coolant Accident (LOCA) (Continued) design profiles unless plant-specific analysis is provided to verify the adequacy of those profiles.
d (2) GE Position:
The test profiles in Appendix A of IEEE 323-1974 shall not be considered acceptable unless plant-specific analysis is provided to justify such profiles.
3.11.2.1.3.1.2 Temperature and Pressure Conditions Inside Containment - Main Steamline Break (MSLB) a.(1) NRC Staf f Position 1. 2. l(1) :
The environmental parameters used for equipment qualification m should be calculated with a plant-specific model reviewed and approved by the staff.
a.(2) GE Position:
See GE Positions 3.11.2.1.3.1.1.a(2), 3. ll . 2.1. 3.1.1. c ( 2 ) ,
and 3.ll.2.1.3.1.1.d(2),
b.(1) NRC Staff Position 1.2.(2):
Models that are acceptable for calculating containment parameters are listed in Subsection 1.1.(2).
b.(2) GE Position:
See GE Position 3.ll.2.1.3.1.1.b(2). _
ggg 3.11-14
GESSAR II 22A7007
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) -
3.11.2.1.3.1.2 Temperature and Pressure Conditions Inside contain-ment - Main Steamline Break (MSLB) (Continued) 3 c.(l) NRC Staff Position 1.2.(3):
In lieu of using the plant-specific containment temperature and pressure design profiles for BWR and ice condenser plants, the generic envelope shown in Appendix C may be used.
4 c.(2) GE Position:
See GE Position 3.11.2.1.3.1.1.c(2).
d (1) NRC Staf f Position 1.2 (4) :
The test profiles included in Appendix A to IEEE Std. 323-1974
() should not be considered an acceptable alternative in lieu of using plant-specific containment temperature and pressure g
M design profiles unless plant-specific analysis is.provided to verify the adequacy of those profiles.
d.(2) GE Position:
See GE Position 3.ll.2.1.3.1.1.d(2).
i e.(l) NRC Staff Position 1.2.(5):
i f
l Where qualification has been completed but only LOCA conditions I were considered, it must be demonstrated that the LOCA qualifi-cation conditions exceed or are equivalent to the maximum calculated MSLB conditions. The following technique is
, acceptable:
i
() e Calculate the peak temperature envelope from an MSLB using a model based on the staff's approved assumptions defined in Subsection 1.1.(2). _
3.11-15 l
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3.11.2.1.3.1.2 Temperature and Pressure Conditions Inside Contain-ment - Main Steamline Break (MSLB) (Continued) e Show that the peak surface temperature of the component to be qualified does not exceed the LOCA qualification temperature by the method discussed in Item 2 of Appendix B.
e If the calculated surface temperature exceeds the qualification temperature, the staff requires that (1) requalification testing be performed with appro-priate margins, or (2) qualified physical protection be provided to assure that the surface temperature will not exceed the actual qualification temperature.
For plants that are currently being reviewed or will be submitted for an operating license review within m six months from issue date of this report, compliance with items (1) or (2) above may represent a substantial impact. For those plants, the staff will consider additional information submitted by the applicant to demonstrate that the equipment can maintain its func-tional operability if its surface temperature rises to the value calculated.
c.(2) GE Position:
Where qualification has been completed but only LOCA conditions were considered, it shall be demonstrated that the LOCA condi-tions exceed or are equivalent to the maximum calculated HELB conditions, or that the critical component of the product being qualified will not be exposed to conditions more severe than those for LOCA. The model used to calculate the peak MSLB temperature envelope shall be based on those assumptions defined in GE Position 3.ll.2.1.3.1.1.b(2).
3.11-16
l GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O ~
3.11.2.1.3.1.3 Effects of Chemical Spray a.(1) NRC Staff Position 1.3:
The effects of caustic spray should be addressed for the equipment qualification. The concentration of caustics used for qualification should be equivalent to or more severe than those used in the plant containment spray system. If the chemical composition of the caustic spray can be affected by equipment malfunctions, the most severe caustic spray environ-ment that results from a single failure in the spray system should be assumed. See SRP Subsection 6.5.2 (NUREG-75/087),
Paragraph II, Item (c) for caustic spray solution guidelines, a.(2) GE Position:
() The effects of spray, where applicable, shall be addressed for g.
g all product qualification. The chemical composition and concentration used for qualification shall be equivalent to or more severe than that used in the containment spray system, taking into account the results of a single failure of the spray system. Water spray caused by activation of fire protection system is not addressed as part of a spray environment.
3.11.2.1.3.1.4 Radiation Conditions Inside and Outside Containment a.(1) NRF Staff Position 1.4:
The radiation environment for qualification of equipment should be based on the normally expected radiation environ-ment over the equipment qualified life, plus that associated
() with'the most severe design basis accident (DBA) during or following which that equipment must remain functional. It ,
3.11-17
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
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0 3.11.2.1.3.1.4 Radiation Conditions Inside and Outside Containment (Continaed) should be assumed that the DBA-related environmental condi-tions occur at the end of the equipment qualified life.
The sample calculations in Appendix D and the following posi-tions provide an acceptable approach for establishing radiation limits for qualification. Additional radiation margins iden-tified in Subsection 6.3.1.5 of IEEE Std. 323-1974 for qualifi-cation type testing are not required if these methods are used.
a.(2) GE Position:
The radiation environment for qualification of products important to safety shall be based on the normally expected ,ggg radiation environment over the product qualified life, plus m
that associated with the most severe DBA during or following which the equipment must remain functional. It shall be assumed that the DBA related environmental conditions occur at the end of the product qualified life.
Appropriate methods similar in nature and scope to that shown in NUREG-0588, Appendix D, for establishing radiation limits for BWR plants for qualification of products will be developed and justified. Additional radiation margins identified in Subsection 6.3.1.5 of IEEE Std. 323-1974 for qualification type testing are not required if these methods are used.
b.(1) NRC Staf f Position 1.4. (1) :
The source term to be used in determining the radiation environment associated with the design basis LOCA should ggg be taken as an instantaneous release from the fuel to the ,
3.11- 18
a
. GESSAR II 22A7707 238 NUCLEAR ISLAND Rev. 5 Or 3.11.2.1.3.1.4 Radiation Conditions.Inside and Outside -
Containment (Continued) atmosphere of 100 percent of the noble gases, 50 percent of
- the iodines, and 1 percent of the remaining fission products.
For all other non-LOCA design basis accident conditions, a
]
source term involving an instantaneous release from the fuel to the atmosphere of 10 percent of the noble gases (except Kr 85, for which a release of 30 percent should be assumed) and 10 percent of the iodines is acceptable.
b.(2) GE Position:
i See Table 3.11-10.
j i c.(1) NRC Staf f Position 1.4. (2) :
The calculation of the radiation environment associated with design basis accidents should take into account the time- o dependent transport of released fission products within 9 various regions of containment and auxiliary structures.
l
, c. (2) GE Position:
j The calculation of the radiation environment associated with a DBA shall take into account the time-dependent transport of released fission products within various regions of containment 1
auxiliary structures.
- d. (1) NRC Staf f Position 1.4. (3) :
The initial distribution of activity within the containment ~
should be based on a mechanistically rational assumption.
Hence, for compartmented containments, such as in a BWR, a large. portion of the source should be assumed to be initially contained in the drywell. The assumption of uniform distribu-tion of activity throughout the containment at time zero is not appropriate. .
3.11-19
GESSAR II 22A7707 238 NUCLEAR ISLAND Rev. 5 3.11.2.1.3.1.4 Radiation Conditions Inside and Outside Containment (Continued)
- d. (2) GE Position:
See Table 3.11-10.
e.(1) NRC Staf f Position 1. 4. (4) :
Effects of engineered safety features (ESP) systems, such as containment sprays and containment ventilation and filtration systems, which act to remove airborne activity and redistribute activity within containment, should be calculated using the same assumptions used in the calculation of of fsite dose.
See SRP Subsection 15.6.5 (NUREG-75/087) and the related sections referenced in the Appendices to that section.
e.(2) GE Position:
m M
The ef fects of the ESF systems, such as containment spray and M containment ventilation and filtration systems, which act to remove activity within the containment, shall be calculated on a basis consistent with the assumptions used to calculate the offsite dose. NUREG-75/087 Subsection 15.6.5 and its appropriate appendices shall be used as guidance.
f.(1) NRC Staf f Position 1.4. (5) :
Natural deposition (i.e., plateout) of airborne activity should be determined using a mechanistic model and best estimates for the model parameters. The assumption of 50 percent in-stantaneous plateout of the iodine released from the core should not be made. Removal of iodine from surfaces by steam condensate flow or washoff by the containment spray may be assumed if such effects can be justified and quantified by analysis or experiment. _
h 3.11-20
GESSAR II 22A7707 238 NUCLEAR ISLAND Rev. 5
() 3.11.2.1.3.1.4 Radiation Conditions Inside and Outside Containment (Continued) f.(2) GE Position:
See Table 3.11-10.
g.(1) NRC Staff Position 1.4.(6):
For unshielded equipment located in the containment, the j gamma dose and dose rate should be equal to the dose and dose rate at the conter point of the containment, plus the nontri-i bution from location-dependent sources, such as the suup water and plateout, unless it can be shown by analyses that location and shielding of the equipment reduce the dose and dose rate.
g.(2) GE Position:
1
()
See Table 3.11-10. p A
h.(1) NRC Staf f Position 1.4. (7) :
I For unshielded equipment, the beta doses at the surface of the
]
equipment should be the sum of the airborne and plateout sources. The airborne beta dose should be taken as the beta dose calculated for a point at the containment center.
f l h.(2) GE Position:
l
, See Table 3.11-10.
i.(1) NRC Staf f Position 1. 4. (8) :
Shielded components need be qualified only to the gamma radia-tion levels required, provided an analysis or test shows that
() the sensitive portions of the ccmponent or equipment are not exposed to beta radiation or that the effects of beta radiation i
I 3.11-21 i
w -+e--+-rvy ypyy-w.~-g m-rg- es g- - -
W--y e-.-q -,ia- yv~-u-mrem
GESSAR II 22A7707 238 NUCLEAR ISLAND Revo 5
~
3.11.2.1.3.1.4 Radiation Conditions Inside and Outside Containment (Continued) llh heating and ionization have no deleterious effects on component performance.
i.(2) GE Position:
When analysis or test data are provided that show that the sensitive portions of a component or product are not exposed to beta radiation or beta radiation heating and ionization effects, or that these effects are not deleterious, then the shielded components shall be qualified only to the gamma radia-tion levels as determined by the mJthods described in GE Position 3.ll.2.1.3.1.4.f.(2).
- j. (1) NRC Staf f Position 1.4. (9) :
Cables arranged in cable trays in the containment should be p ll) assumed to be exposed to half the beta radiation dose calculated M for a point at the center of the containment, plus the gamma ray dose calculated in accordance with Subsection 1.4(6). This reduction in beta dose is allowed because of the localized shielding by other cables, plus the cable tray itself.
j.(2) GE Position:
See Table 3.11-10.
K.(1) NRC Staf f Position 1. 4. (10) :
Paints and coatings should be assumed to be exposed to both beta and gamma rays in the assessment of their resistance to radiation. Plateout activity should be assumed to remain on the equipment surface unless the effects of the removal mechan-isms, such as spray washoff or steam condensate flow, can be ll) justified and quantified by analysis or experiment. _
l 3.11-22
GESSAR II 22A7707 238 NUCLEAR ISLAND Rev. 5
() 3.11.2.1.3.1.4 Radiation Conditions Inside and Outside Containment (Continued)
~
k.(2) GE Position:
i Paints and coatings shall be assumed to be exposed to both beta and gamma rays in the assessment of their resistance to radiation. Platcout activity shall be assumed to remain on g the equipment surface unless the effects of the removal mechan-j isms, such as spray washoff or steam condensate flow, can be '
justified and quantified by analysis or experiment.
- 1. (1) NRC Staf f Position 1.4. l(11) :
Components of the emergency core cooling system (ECCS) located outside containment (e.g., pumps, valves, seals, and electrical equipment) should be qualified to withstand the radiation equivalent to that penetrating the containment, plus the
(} exposure from the sump fluid using assumptions consistent with the requirements stated in Appendix K to 10CFR50.
p 4
1.(2) GE Position:
i See Table 3.11-10.
1 l~ m. (1) NRC Staf f Position 1.4. (12) :
4 Equipment that may be exposed to radiation doses below 10 rads should not be considered exempt from radiation qualification unless analysis supported by test data is provided to verify that these levels will not degrade the operability of the f
equipment below acceptable values.
m.(2) GE Position:
(} Products that may be exposed to radiation doses at or below 10 rads shall not be considered exempt from radiation 3.11-23
+ , , - . - - _ .,_.-4 , - - - , , . - .,, ._ . . - . - , , y. _ , ,,,m_m. ,_. _, , -,_
l GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
~
3.11.2.1.3.1.4 Radiation Conditions Inside and Outside Containment (Continued) qualification unless analysis supported by test data is pro-vided to verify that these levels will not degrade the opera-bility of the product below acceptable values.
- n. (1) NRC Staf f Position 1.4. (13) :
The staff wiil accept a given component as qualified, provided it can be shown that the component has been qualified to inte-grated beta and gamma doses which are equal to or higher than those levels resulting from an analysis similar in nature and scope to that included in Appendix D (which uses the source term given in item (1) above), and that the component incorpor-ates appropriate factors pertinent to the plant design and operating characteristics, as given in these general guidelines.
e GE Position: m n.(2) m' An analysis similar in nature and scope to that in NUREG-0588, Appendix D, shall be developed for each product type and application which incorporates the appropriate factors perti-nent to the plant design and operating characteristics.
- o. (l) NRC Staf f Position 1.4. (14) :
When a conservative analysis has not been provided by the applicant for staff review, the staff will use the radiation environment guidelines contained in Appendix D, suitably corrected for the differences in reactor power level, type, containment size, and other appropriate factors.
o.(2) GE Position:
O See Table 3.11-10. ,
3.11-24
i l
i GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
() ~
3.11.2.1.3.1.5 Environmental Conditions for Outside Primary Containment
- a. (1) NRC Staf f Position 1. 5. (1) :
Equipment located outside containment that could be subjected i to high energy pipe breaks should be qualified to the conditions
< resulting from the accident for the duration required. The techniques for calculating the environmental parameters described in Subsections 1.1 through 1.4 above should be 4 applied.
a.(2) GE Position:
t Products important to safety located outside primary contain-ment that could be subjected to high energy pipe breaks shall be qualified to the conditions resulting from the accident $
for the duration required. The techniques for calculating M the resulting environmental parameters deceribed in GE Position l Statements 3.11.2.1.3.1.1 through 3.11.2.1.3.1.4 shall be applied.
f j b. (1) NRC Staf f Position 1. 5. (2) :
' Equipment located in general plant areas outside containment l
where equipment is not subjected to a design basis acci. dent
- environment should be qualified to the normal and abnormal range of environmental conditions postulated to occur at the equipment location.
b.(2)' GE Position: r Products important to safety, located in general plant areas outside containment where the product is not subjected to a ,
1 3.11-25
,,e , . . ..r - ,, r e,-. . . _ , , - - - - , , - - , -.~,-,----.,--.----n-. - , - ,,- , ,
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 3.11.2.1.3.1.5 En/ironmental Conditions for Outside Primary 9
Containment (Continued)
DBA environment., shall be qualified to the normal and abnormal range of envircnmental conditions expected to occur at the product locatio1.
- c. (1) NRC Staff Position 1.5.(3):
Equipment not served by Class lE environmental support systems or served by Class lE support systems that may be secured during plant operaticn or shutdown should be qualified to the limiting environmental conditions that are postulated for that location, assuming a loss of the environmental support system.
c.(2) GE Position:
See Table 3.11-10.
3.11.2.1.3.2 Qualification Methods 3.11.2.1.3.2.1 Selection of Methods
- a. (1) NRC Staf f Position 2.1. (1) :
Qualification methods should conform to the requirements of IEEE 323-1974.
a.(2) GE Position:
Qualification methods shall conform to the criteria defined in Subsection 3.11.2.2.4. ,
O 3.11-26
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O 3.11.2.1.3.2.1 Selection of Methods (Continued) -
b.(1) NRC Staf f Position 2.1, (2) :
The choice of the methods is largely a matter of technical judgment and availability of information that supports the conclusions reached. Experience has shown that qualification of equipment subjected to an accident environment without test data is not adequate to demonstrate functional operability.
In general, the staff will not accept analysis in lieu of test data unlesst (1) testing of the component is impractical be-cause of size limitations, and (2) partial type test data are provided to support the analytical assumptions and conclusions reached.
b.(2) GE Position:
See Table 3.11-10.
m c.(1) NRC Staff Position 2.1.(3):
The environmental qualification of equipment exposed to DBA environments should conform to the following positions. The bases should be provided for the time interval required for operability of this equipment. The operability and failure criteria should be specified and the safety margins defined.
e Equipment that must function in order to mitigate any accident should be qualified by test to demon-strate its operability for the time required in the environmental conditions resulting from that accident.
e Any equipment (safety-related or nonsafety-related) that need not function in order to mitigate any acci-dent but that must not fail in a manner detrimental to plant safety should be qualified by test to 3.11-27
GESSAR II 22A7007 230 MUCLEAR ISLAND Rev. 5 3.11.2.1.3.2.1 Selection of Methods (Continued) demonstrate its capability to withstand any accident environment for the time during which it must not fail, o Equipment that need not function in order to mitigate any accident and whose failure in any mode, in any accident environment, is not detrimental to plant safety, need be qualified only for its nonaccident service environment. Although actual type testing is preferred, other methods when justified may be found acceptable. The bases should be provided for concluding that such equipment is not required to function in order to mitigate any accident and that its failure in any mode, in any accident environment, is not detrimental to plant safety.
GE Position:
O c.(2) m The environmental qualification of products exposed to DBE (LOCA or HELB) environments shall conform to the following positions. The cognizant engineer establishes the bases associated with the time interval required for operability of this product. The operability and acceptance criteria shall be specified and the safety margins defined.
e Products that must function in order to mitigate an accident shall be qualified to demonstrate operability for the time required in the environmental conditions resulting from that accident. Type testing shall be the preferred method of qualification, subject to the limitations described in paragraph
- 3. ll . 2.1. 3. 2.1. b . ( 2 ) above. .
O 3.11-28
-- -- . ~. .- .- . .- _ _.
i GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
.I
~
3.11.2.1.3.2.1 Selection of Methods (Continued) e Any product that is not required to. function to miti-l gate an accident but that must not fail in a manner detrimental to plant safety or provide false informa-l' tion to an operator shall be qualified to demonstrate its capability to withstand any accident environment for the time during which it must not fail. Type testing shall be the preferred method of qualification, subject to the limitations described in paragraph 3.11.2.1.3.2.1.b.(2) above.
e Any product'that is not required to function in order to mitigate any accident and whose failure in any mode
! in any accident environment is not detrimental to plant safety is considered not important to safety, non-essential, and exempt from qualification because failure would be of no consequence either to plant safety or to the health and safety of the public. -
- d. (1) NRC Staff Position 2.1.(4):
i For environmental qualification of equipment subject to events,
$ other than a DBA, which result in abnormal environmental con-ditions, actual type testing is preferred. However, analysis or operating history, or any applicable combination thereof, coupled with partial type test data, may be found acceptable, subject to the applicability and detail of information pro-vided.
3.11.2.1.3.2.2 Qualification by Test
- a. (1) NRC Staf f Position 2.2. (1) :
The failure criteria should be established prior to testing.
(~/)
s.
i 3.11-29 i
~ , - - ., _,-- -,- ,- .
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 3.11.2.1.3.2.2 Qualification by Test (Continued)
- a. (2) GE Position:
See Table 3.11-10.
- b. (1) NRC Staf f Position 2.2. (2) :
Test results should demonstrate that the equipment can per-form its required function for all service conditions postu-lated (with margin) during its installed life.
- b. (2) GE Position:
Test results shall demonstrate that the product can perform its design function for all service conditions postulated (with margin) during its installed life.
mg
- c. (1) NRC Staf f Position 2.2. (3) 9W m
The items described in Subsection 6.3 of IEEE Std. 323-1974, supplemented by Items (4) through (12) (NRC Staff Positions 2.2(4) through 2.2(12)) below, constitute acceptable guidelines for establishing test procedures.
- c. (2) GF Position:
Section 4 of this document defines the acceptable guidelines for establishing test procedures. These procedures are con-sistent with subsection 6.3 of IEEE Std. 323-1974.
- d. (l) NRC Staf f Position 2.2. (4) :
When establishing the simulated environmental profile for qualifying equipment located inside containment, it is pre-ferred that a single profile be used that envelops the .
3.11-30
GESSAR II 22A7007 t
I 238 NUCLEAR ISLAND Rev. 5 i
~
i 3.11.2.1.3.2.2 Qualification by Test (Continued) environmental conditions resulting from any design basis l
event during any mode of plant operation (e.g., a profile l
that envelops the conditions produced by the main steamline i break and loss-of-coolant accidents).
d.(2) GE Position:
When establishing the simulated environmental profile for qualifying the product located inside containment, a singic profile may be used that envelops the environmental conditions resulting from any design basis event (DBE) during any mode of plant operation for which the product is required to function. ,
c .
t e.(1) NRC Staff Position 2.2.(5):
1 e
! (:) Equipment should be located above flood level or protected m
against submergence by location in qualified watertight enclosures. Where equipment is located in watertight enclosure, qualification by test or analysis should be used l
j to demonstrate the adequacy of such protection. Where j equipment could be submerged, it should be identified and demonstratedtobequalifiedbhtestforthedurationrequired.
e.(2) GE Position:
Products should be located above flood level or protected against submergence by location in watertight enclosures.
Where a product is located in a watertight enclosure, quali-fication by test or analysis shall be used to demonstrate the adequacy of such protection. Products not located in watertight enclosures, that could be subject to submergence through which they must remain functional, shall be demon-(} strated to be qualified-by test for the duration required.
, 3.'11-31
(
o c , . . ,,,--,,,y
._...__y_- , . . . - -r --r- --e---mv---w-s- r-e av w - w --w--, **-*-=p-'w-~C "* t e' e w <-----"==8e- -*--
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.+ .
'G5SSAR 11 -
'/ 22A7007 238-NUCLEAR ISLAND .- -
. Rev. 5 a
1 3.11.2.1.3.2.2 Qualification by Test (Continued) f.(1) NRC Staf f Positicb 2. 2. (6) : ,
The temperature to which, equipment is qualifie'd, wnen exposed to the simulated accident enviropment, should be defined by thermcouple readings on or as close as practical)to the sur- ,
i '
face of the component being qualified. ,
I
~
f.(2) GE Position: ,..'i s. ,-
r
+
'_ a Cue Table 3.11--10. ,,
r
"^
- g. (1) NRC Staff Position 2.2.(i}: '
s , ,-
/ .' r," ,
Performance characteristics of equipment should be' verified before, after, and: periodically during testing throughout its range of required operability. ,
,4
< ,r' i' .
4, mh O
m g.(2) GE Position: -
j . .
/
Performance characteristics of the product shall. be verified before, periodically during, and after testing throughout its range of required oporabili.ty defined in the product performance specification. ,
- r. ,
l .. ,
h.(1) NRC S ta f f Positi.o" 2. 2. ( 8) : ,
Caustic spray should be incorporated during simulated event testing at-the maximum pres,sure and at the temperature condi-tions that.would occur whe( the,:onsite' spray systems actuate. ,. a o i h.(2) GE Position: ;
r ,
.a
,? .
See Table 3.11-10. > i
_ ,O s -
./
f 3.112 32 i
, GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
~
3.11.2.1.3.2.2 Qualification by Test (Continued) i.(1) NRC Staf f Pocition 2.2. (9) :
The operability status of equipment should be monitored continuously during testing. For long-term testing, however, monitoring at discrete intervals should be justified if used.
l- i. (2) GE Position:
, f; / ,/' ,
,See Table 3.11-10.
{#
0
- j . (M) NRC Staf f Position 2.2. (10) :
o p -
Expected extremes in power supply voltage range and frequency 7 ,shoulp be applied during simulated event environmental testing.
i pf ;'<
/
jl(2) GE Position: m
}
.v*
+ e i .Expkctedextremesinpowersupplyvoltage, range, and frequency, as defined in the product performance specification, shall be anplied during simulated event environmental testing.
, r 1
k.Yl) NRC Staf f Position 2.2. (11) :
e s Dust) environments should be addressed when establishing quali-
- i. 7-fication service conditions.
- k. (2) GE. Position:
w See Tablo 3.11-10.
7_ _
- 1. (1) ,'URC Staf f Position 2.2. (12) :
/~T 'l Cobalt-60 is an acceptable gamma radiation source for
(_/
environmental qualification.
l -
3.11-33
.f 4;
I l 5
. ~ - _
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 3.11.2.1.3.2.2 Qualification by Test (Continued) 1.(2) GE Position:
Cobalt-60 is a commonly used and acceptable gamma radiation source for the product qualification. Other sources capable of providing appropriate energy level for irradiation of a test sample (e.g., Cesium-137) are also acceptable.
3.11.2.1.3.2.3 Test Sequence
- a. NRC Staf f Position 2. 3. (1) :
The test sequence should conform fully to the guidelines established in Subsection 6.3.2 of IEEE Std. 323-1974. The test procedure should ensure that the same piece of equipment is used throughout the test sequence and that the test simu-lates as closely as practicable the postulated accident environment.
- b. GE Position:
The test sequence to be followed by GE is defined in Subsection 3.11.2.2.4. The test procedures shall ensure that the same test specimen is used throughout the test sequence and that the test simulates as closely as practicable the postulated acci-dent environment.
3.11.2.1.3.2.4 Other Qualification Methods
- a. NRC Staff Position 2.4:
Qualification by analysis or operating experience implemented, as described in IEEE Std. 323-1974 and other ancillary stan-dards, may be found acceptable. The adequacy of these methods will be evaluated on the basis of the quality and detail of .
3.11- 34
GESSAR II 22A7007 238 NUCLEAR ISLAND
. O 3.11.2.1.3.2.4 Other Qualification Methods (Continued) of the information submitted in support of the assumptions made and the specific function and location of the equipment.
These methods are most suitable for equipment where testing is precluded by physical size of the equipment being qualified.
It is required that, when these methods are employed, some partial type tests on vital components of the equipment be provided in support of these methods.
- b. CE Position:
Qualification by analysis or operating experience is defined in Subsection 3.11.2.2.4.
3.11.2.1.3.3 Margins m
M
() a. (1) NRC Staf f Position 3.0. (1) :
Quantified margins should be applied to the design parameters i discussed in Section 1 to assure that the postulated accident conditions have been enveloped during testing. These margins should be applied to any margins (conservatism) applied during the derivation of the specified plant parameters.
a.(2) GE Position:
I i Quantified test margins shall be applied to the design environ-mental parameters, or it shall be shown that adequate margin is already included in the environmental requirements. In either
! case, the margins shall be justified as adequate and i
i documented. .
g
' 3.11-35 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 ,
l 3.11.2.1.3.3 Margins (Continued)
- b. (1) NRC Staf f Position 3. (2) :
In lieu of other proposed margins that may be found acceptable, the suggested values indicated in IEEE Std. 323-1974, Sub-section 6.3.1.5, should be used as a guide. (Note exceptions stated in Subsection 1.4.)
b.(2) GE Position:
Qualification test margins shall be determined in accordance with the criteria presented in Subsection 3.11.2.2.4.
c.(1) NRC Staff Position 3.(3):
When the qualification envelope in Appendix C is used, the only required margins are those accounting for the inaccuracies ,
in the test equipment. Sufficient conservatism has already 9 m
been included to account for uncertainties such as production errors associated with defining satisfactory performance (e.g.,
when only a small number of units are tested).
c.(2) GE Position:
The qualification environmental profiles identified in Appendix C of NUREG-0588 may be used for in-containment equipment with only margin added for inaccuracies of the test equipment.
d.(l) NRC Staf f Position 3. (4) :
Some equipment may be required by the design to perform its safety function within only a short time period into the event (i.e., within seconds or minutes), and, once its func-tion is complete, only when subsequent failures are shown not to be detrimental to plant safety. Other equipment may not 3.11-36
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
~
3.11.2.1.3.3 Margins (Continued) be required to perform a safety function but must not fail within a short time period into the event and when. subsequent failures are also shown not to be detrimental to plant safety.
Equipment in these categories is required to remain functional in the accident environment for a period of at least one hour in excess of the time assumed in the accident analysis. For all other equipment (e.g., postaccident monitoring, recombiners, etc.) the 10 percent time margin identified in Subsection 6.3.1.5 of IEEE Std. 323-1974 may be used.
d.(2) GE nosition:
See Table 3.11-10.
j 3.11.2.1.3.4 Aging m
O
\/ $
- a. (1) NRC Staf f Position 4. (1) : m i
Aging effects on all equipment, regardless of its location in the plant, should be considered and included in the qualifi-cation program.
a.(2) GE Position:
Aging' effects on all products important to safety, regardless of their location in the plant, shall be considered and addressed in the qualification program.
- b. (1) NRC Staff Position 4.(2):
The degrading influences discussed in Subsections 6.3.3, 6.3.4,
, and 6.3.5 of IEEE Std. 323-1974 and the electrical and mechanical stresses associated with cyclic operation of
} equipment should be considered and included as part of the aging programs. "
3.11-37
GESSAR Il 22A7007 238 NUCLEAR ISLAND Rev. 5
~
3.11.2.1.3.4 Aging (Continued)
- b. (2) GE Position:
The degrading influences discussed in Subsection 3.11.2.2.4 and the electrical and mechanical stresses associated with cyclic operation of the product shall be considered and addressed as part of the aging programs.
- c. (1) NRC St'aff Position 4.(3):
Synergistic effects should be considered in the accelerated ,
aging programs. An engineering evaluation shall be performed m to assure that no known synergistic effects have been identi-fied on materials that are included in the equipment being qualified. Where synergistic effects have been identified, they should be accounted for in the qualification programs.
Refer to NUREG/CR-0276 (SAND 78-0799) and NUREG/CR-0401 (SAND 78-1452), Qualification Testing Evaluation Quarterly Reports, for additional information.
c.(2) GE Position:
~
See Table 3.11-10.
O 3.11-38
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
-3.11.2.1.3.4
. Aging (Continued)
- d. (1) NRC Staf f Position 4. (4) :
The Arrhenius methodology is considered an acceptable method of addressing accelerated aging. Other aging methods that can be supported by type tests will be evaluated on a case-by-case basis.
- d. (2) GE Position:
The Arrhenius methodology is considered an acceptable method of addressing accelerated thermal aging. Other aging methods may be considered appropriate if justified for the application. l
- e. (l) NRC Staff Position 4. (5) :
Known material phase changes and reactions-should be defined ,
J to ensure that no known changes occur within the extrapolation.
limits,
]
i e.(2) GE Position:
See Table 3.11-10.
f.(1) NRC Staff Position 4. (6) :
The aging acceleration rate used during qualification testing and the basis upon which the rate was established should be described and justified.
f.(2) GE Position:
The aging acceleration rate used during qualification testing
- and the basis upon which the rate was established shall be
' described and justified. .
3.11-39
GESSAR II 22A7007 238 MUCLEAR ISLAND Rev. 5 3.11.2.1.3.4 Aging (Continued)
- g. (1) NRC Staf f Position 4. (7) :
Periodic surveillance testing under normal service conditions is not considered an acceptable method for ongoing qualification unless the plant design includes provisions for subjecting the equipment to the limiting service environment conditions (specified in Subsection 3(7) of IEEE Std. 279-1971) during such testing.
- g. (2) GE Position:
Periodic surveillance testing under normal service conditions shall not constitute an acceptable method of ongoing qualifi-cation unless provisions for periodically confirming the product performance capability under the limiting service m environmental conditions are included in the ongoing qualifi-m cation program.
- h. (1) NRC Staff Position 4.(8):
Effects of relative humidity need not be considered in the aging of electrical cable insulation.
h.(2) GE Position:
See Table 3.11-10.
- i. (1) NRC Staf f Position 4. (9) :
The qualified life of the equipment (and/or component, as applicable) and the basis for its selection should be defined. a O
3.11-40
}
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 3.11.2.1.3.4 Aging - (Continued) i.(2) GE Position:
The qualified life of the product shall be defined on the basis of the aging program conducted.
j . (1) NRC Staf f Position 4. (10) :
Qualified life should be established on the basis of the severity of the testing performed, the conservatisms employed in the extrapolation of data, the operating history, and in other methods that may be reasonably assumed, coupled with good engineering judgment.
j.(2) GE Position:
(} The qualified life shall be established on the basis of the aging program and shall consider the severity of the testing
]
performed, the conservatisms employed in the extrapolation of data, the operating history, or other methods that can be reasonably assumed, coupled with good engineering judgment.
3.11.2.1.3.5 Qualification Documentation
- a. (1) NRC Staf f Position 5. (1) :
The. staff endorses the requirements, stated in IEEE Std. 323-1974, that "the qualification documentation shall verify that each type of electrical equipment is qualified for its application and meets its specified performance requirements.
The basis of qualification shall be explained to show the relationship of all facets of proof needed to support adequacy of the complete equipment. Data used to demonstrate the qualification of the equipment shall be pertinent to the I
(} application and organized in an auditable form." -
3.11-41
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
~
3.11.2.1.3.5 Qualification Documentation (Continued)
- a. (2) GE Position:
Qualification documentation shall conform to the commitments made in Subsection 3.11.2.2.4.
- b. (1) NRC Staff Position 5.(2):
The guidelines for documentation in IEEE Std. 323-1974, when fully implemented, are acceptable. The documentation should ,
include sufficient information to address the required m
information identified in Appendix E. A certificate of conformance by itself is not acceptable unless it is accompanied by test data and information on the qualification program.
b.(2) GE Position:
Qualification documentation and files shall conform to the commitments made in Subsection 3.11.2.2.4. -
3.11.2.2 Class lE Product Environmental Qualification Basis 3.11.2.2.1 Scope This subsection provides a compilation of requirements for the environmental qualification of Class lE products
- within the Nuclear Isand scope of responsibility. It is further restricted to those Class lE products designed for applications which are exposed to environments resulting from design basis events (except for those which will experience only dynamic events) .
- " Product" used herein is synonymous with " Class lE Product". _
3.11-42
1 GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 1
_/
3.11.2.2.1 Scope (Continued)
The manufacturers and users of Class lE products are required to provide assurance that each product will meet or exceed its performance requirements throughout its installed life. This is accomplished through a disciplined program of quality assurance that includes but is not limited to design, qualification, production quality control, installation, maintenance, and periodic testing.
It is the primary role of qualification to assure that, for each type of Class lE product, the design and the manufacturing pro-cesses are such that there is a high degree of confidence that the product will perform as required in the specified environment.
Other steps in the quality assurance program invoke strict design and manufacturing control to assure that all products of the same type match that which was qualified and are suitabley applied,
~
installed, maintained, and periodically tested.
(%)'
(
The methods described in this subsection apply to work accomplished by the General Electric Nuclear Energy Business Operation (NEBO) as well as all its vendors and contractors.
3.11.2.2.2 Applicable Documents If a conflict exists between the requirements contained in this subsection and those in a listed document, those in this sub-section shall govern.
3.11.2.2.2.1 General Electric Documents
, The following documents form a part of these requirements.
, (1) Design-Record Files O
i 3.11-43 I
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O
3.11.2.2.2.1 , General Electric Documents (Continued)
(2) EMI Susceptibility Test Specification (3) Regulatory Guide Implementation Positions 3.11.2.2.2.2 Codes, Standards, and Regulations The following codes, standards, and regulations, as interpreted by NEBO, form a part of these requirements to the limits specified in this subsection:
(1) NRC Regulatory Guides
- No. Title (a) 1.40 Qualification Tests of Continuous Duty Motors Installed Inside the Containment of Water-Cooled Nuclear Power Plants (generally accepts IEEE 334-1971)
(b) 1.63 Electric Penetration Assemblies in Con-tainment Structures for Water-Cooled Nuclear Power Plants (generally accepts IEEE 317-1976)
(c) 1.73 Qualification Tests of Electric Valve Operators Installed Inside the Contain-ment of Nuclear Power Plants (generally accepts IEEE 382-1972)
- See Section 1.8 for revision and date. .
3.11-44
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O 3.11.2.2.2.2 Codes, Standards, and Regulations (Continued)
~
No. Title (d) 1.89 Qualification of Class lE Equipment for Nuclear Power Plants (generally accepts IEEE 323-1974)
(e) 1.92 Combining Modal Responses and Spatial Components in Seismic Response Analysis (f) 1.100 Seismic Qualification of Electric Equipment for Nuclear Power Plants (generally accepts IEEE 344-1975)
(g) 1.131 Qualification Tests of Electrical Cables, Field Splices, and Connections for Light-gg Water-Cooled Power Plants (generally
! \/ accepts IEEE 383-1974) l (2) American National Standards Institute (ANSI)
(a) ANSI N45.2 - Quality Assurance Program Requirements for Nuclear Power Plants (b) ANSI N45.2.ll - Quality Assurance Requirements for the
, Design of Nuclear Power Plants i (c) ANSI N45.2.13 - Quality Assurance Requirements for Control of Procurement of Items and Services for l Nuclear Power Plants (d) ANSI B31.1 - Power Piping
~
gS (3) Institute of Electrical and Electronic Engineers, Inc.
k/ (IEEE) 3.11-45
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
~
3.11.2.2.2.2 Codes, Standards, and Regulations. (Continued)
(a) IEEE 627-1980 - IEEE Standard for Design Qualification of Safety Systems Equipment Used in Nuclear Power Generating Stations (b) IEEE 323-1974 - IEEE Standard for Qualifying Class lE Equipment for Nuclear Power Generating Stations (c) IEEE 344-1975 - IEEE Recommended Practices for Seismic Qualification of Class lE Equipment for Nuclear Power Generation Stations (d) IEEE 381-1977 - IEEE Standard Criteria for Type Tests of Class lE Modules Used in Nuclear Power Generating Stations (e) IEEE 317-1976 - IEEE Standard for Electric Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations (f) IEEE 334-1974 - IEEE Standard for Type Tests of Con-tinuous Duty Class IE Motors for Nuclear Power Generating Stations (g) IEEE 382-1980 - IEEE Standard for Qualification of Safety-Related Valve Actuators (h) IEEE 383-1974 - IEEE Standard for Type Test of Class lE Electric Cables, Field Splices, and Connections for Nuclear Power Generating Stations (i) IEEE 649-1980 - TEEE Standard for Qualifying Class lE Motor Control Centers for Nuclear Power Generating Stations _
3.11-46
GESSAR II 22A7007 238 NUCLEAR JSLAND Rev. 5 3.11.2.2.2.2 Codes, Standards, and Fegulations (Continued)
(j) IEEE 650-1979 - IEEE Standard for Qualification of Class lE Static Battery Chargers and Inverters for t
Nuclear Power Generating Stations (k) IEEE 634-1978 - IEEE Star.dard Cable Penetration Fire Stop Qualification Test (1) IEEE 501-1978 (ANSI C37.98) - IEEE Standard Seismic Testing of Relays (m) IEEE 535-1979 - IEEE Standard for Qualification of Class lE Lead Storage Batteries for Nuclear Power Generating Stations (4) American Society of Mechanical Engineers (ASME)
~
O ASME Boiler and Pressure Vessel Code Sections II, III, (a)
IV, V, VIII, and XI 3.11.2.2.3 Reference Documents (1) IEEE 101, IEEE Guide for the Statisitcal Analysis of Thermal Life Test Data, The Institute of Electrical and Electronics Engineers, Inc. - 1972.
(2) EPRI NP-1558, A Review of Equipment Aging Theory and Technology, Electric Power Research Institute, 4
September 1980.
(3) NUREG/CR-0275 (SAND 78-0067), An Experimental Investi-gation of Synergisms in Class 1 Components Subjected to LOCA Type Tests, Sandia Laboratories, August 1978.
f)T
~.
3.11-47
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O
3.11.2.2.3 Reference Documents (Continued)
(4) NUREG/CR-0401 (SAND 78-1452) and NUREG/CR-0276 (SAND 78-0799), Qualification Testing Evaluation Quarterly Reports, Sandia Laboratories, 1978.
(5) NUREG-0588, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment, December 1979.
(6) NUREG/CR-2156 (SAND 80-2149), Radiation Thermal Degrada-tion of PE and PVC: Mechanism of Synergism and Dose Rate Effects, R. Clough and K. Gillen, Sandia Laboratories, June 1981.
(7) NUREG-0484, Rev. 1.
O O
3.11-48
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 3.11.2.2.4 Program Description PROPRIETARY INFORMATION - provided under separate cover.
1 O l i
i.
3.11-49 through 3.11-81
P 1
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 3.11.3 Qualification Results lh Applicant to supply. _
3.11.4 Loss of Ventilation All environmental zones of tne Reactor and Turbine Buildings and specific environmental zones of the Auxiliary Building (AB-5, AB-6, and AB-7) and Fuel Building (FB-2 and FB-3 of Table 3.11-4) include the loss of heating, ventilating, and air conditioning (HVAC) systems as an abnormal environmental condition. The HVAC system for the Control Building and the environmental zones of the other buildings are designed as Class lE electrical equipment.
Hence, no special loss of HVAC environmental conditions have been included in these environmental zones. The bases ensuring that the loss of HVAC conditions are not exceeded are provided in the following subsections. Testing and documentation of safety-related equipment exposed to loss of HVAC environmental condi-lll tions is covered under Subsection 3.11.3. _
3.11.4.1 Drywell Loss of HVAC Applicant to supply.
3.11.4.2 Containment Loss of HVAC Applicant to supply, o
w 3.11.4.3 Auxiliary Building Loss of HVAC Applicant to supply.
3.11.4.4 Fuel Building Loss of HVAC Applicant to supply. ,
3.11-82
-GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
() 3.11.4.5 Turbine Building Loss of HVAC Applicant to supply.
3.11.4.6 Diesel Generator Building Loss of HVAC The Diesel Generator Buildings are served by safety class HVAC o systems. Thus the loss of the HVAC systems for these buildings y is not considered credible, and these environmental zones need not be analyzed for a loss of HVAC event. Also, since division-ally separate diesel generation systems are provided with their dedicated, divisionally separate HVAC systems, the loss of one HVAC system will not adversely affect the operability of safety-related equipment located in these buildings. _
3.11.5 Estimated Chemical and Radiation Environment Applicant to supply.
(]} .
O 3.11-83/3.11-84
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 5 Table 3.11-1 ENVIRONMENTAL ZONES Zone Description . Typical Equipment Drywell*
DW-1 Outside RPV shield Recirculation System equipment ,
wall - not at core Recirculation suspension midplane DW-2 Outside RPV shield Condenser chambers ,
wall - at core Gross gamma detection midplane Safety / relief valves MSIVs DW-3 Under RPV Control rod drives Neutron monitors Vessel skirt flange
{ }- Bottom head insulation CRD removal platform DW-4 Drywell dome Vessel head and insulation Vessel-to-drywell seal
- CT-1 Above refueling floor Refueling platform l
Fuel preparation machines Radiation and atmosphere monitors
- See Figure 3.11-1 3.11-85 -
l
\ ., ,.- --. - -. - . , - . . . - - _ - ., - - - . . . . . . - . -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 Table 3.11-1 h ENVIRONMENTAL ZONES (Continued)
Zone Description Typical Equipment CT-2 Immediately above TIP Drive System suppression pool Recirculation flow control hydraulic equipment C&I panels CT-3 HCU floor HCU units and panels Panels multiplexer CT-4 SLCS area SLC System equipment and panel CT-5** RWCU rooms RWCU heat exchanger RWCU F/D units and pumps RWCU backwash receiving tank RWCU valve holding pump Auxiliary Building AB-1 Electric switchgear Flow indicating control and remote shutdown Inverters panel area Converters RCIC control equipment Remote shutdown panel IRM and SRM preamplifiers AB-2 LPCS, HPCS, RHR C LPCS, HPCS, RHR C pumps, motors Panels and instruments Leak detection equipment
- Contains high energy piping - failure is postulated 3.11-86 .
.GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 Table 3.11-1 ENVIRONMENTAL ZONES (Continued)
Zone - Description Typical Equipment AB-3** RCIC turbine and pump RCIC pump rooms Turbine, gland seal and compressor Panel leak detection equipment AB-4** RHR A and B pump rooms RHR pumps and motors, heat exchanger panel Leak detection equipment AB-5** RWCU pump rooms RWCU pumps and motors Panel Leak detection equipment O AB-6 Corridors outside ECCS ESW Process Rad. Monitors i
rooms i
AB-7** Steam tunnel MSIVs and NS Shutoff System Leak detection equipment AB-8 Battery rooms Batteries HVAC equipment 4
AB-9 Air Positive Seal APS compressor System room APS air receiver Fuel Building FB-1 Fuel pool pump area Fuel pool pump and motors Radiation monitors Panels Leak detection equipment "T Fuel pool heat exchangers
[G
- Contains high energy piping - failure is postulated 3.11-87
-- .~ ,,e- 9 y- . v --
e, .-y . ------w~w r mv- --- - --- -- - + - - - m- r e
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 Table 3.11-1 ENVIRONMENTAL ZONES (Continued)
Zone Description Typical Equipment FB-2 Operating floor Neutron System amplifiers Pool Radiation monitors Auxiliary platforms Fuel preparation machines Recirculating LF MG sets FB-3 Below operating floor CRD pumps and motors Radiation monitors Leak detection equipment FB-4 SGTS filter rooms SGTS filters FB-5 SGTS fan rooms SGTS Fans gg FB-6 Shield annulus fan Fans rooms Room coolers Control Building i
CB-1 Control and control Control room equipment equipment rooms Control equipment panels Computer CB-2 IIVAC equipment Fans room Hinter chillers Chiller pumps
- Contains high energy piping - failure is postulated lll
- 3.11-88 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O
~
Table 3.11-1 ENVIRONMENTAL ZONES.(Continued)
Zone Description Typical Equipment Diesel Generator Building DG-1 General areas Diesel generator control panel
. Diesel generator motor Control Center Engine generator room Batteries (HPCS)
Switchgear, battery, Switchgear (IIPCS) and 11VAC rooms Turbine Building TB-1 Above and below Bypass valves, control valves turbine pressure transmitters Operating floor Tubine Protection System Condensate and Feedwater System control Offgas treatment and monitoring Radiation detectors b
v 3.11-89 /3.11-90 _
O O O Table 3.11-2
! ENVIRONMENTAL CONDITIONS FOR REACTOR BUILDING EQUIPMENT l
't I
f 3
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1 i
To be provided by Applicant w w i w CD l W
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O O O Table 3.11-3 ENVIRONMENTAL CONDITIONS AND LIMITS FOR AUXILIARY BUILDING EQUIPMENT 4
a ~4 To be provided by Applicant co W
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Table 3.11-4
- ENVIRONMENTAL CONDITIONS AND LIMITS FOR FUEL BUILDING EQUIPMENT 3
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W s-To be provi.ded by AppU par.t.~. $ "
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see Table 3.11-8 for notes ;, C
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j Table 3.11-6 ENVIRONMENTAL CONDITIOUS AND LIMITS FOR DIESEL GENERATOR BUILDING EQUIPMENT i
f n
J l
i To be provided by Applicant "
u
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! 3.37 .
i
O O O Table 3.11-7 ENVIRONMENTAL CONDITIONS AND LIMITS FOR TURBINE BUILDING EQUIPMENT i
To be provided by Applicant "
w w CD 5o om e t* V3 o M 0)
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- 3 See Table 3.11-8 for notes viO
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3.37
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O
1 Table 3.11-8 NOTES FOR TABLES 3.11-2 THROUGH 3.11-7
- t~~
M To be provided by Applicant _
O i
i i
O
.i 3.11-103/3.11-104 n - - - - - , - , - - - - , _ _ - , , . - - - , , ~ . - . . . - ...-- - -- . .--.-- ,---,:,,,.,., ,, .- , - - - . ~ . ,
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 i Table 3.11-9
- SAFETY-RELATED EQUIPMENT IDENTIFICATION '
AND ENVIRONMENTAL QUALIFICATION
SUMMARY
This table provides an identification and environmental qualifica-tion summary for safety-related equipment for each applicable system through a series of tables. Each table is designated as Table 3.11-9 (MPL) where MPL is the master parts list number for the system. Each table lists the items for a particular system by MPL item number and make, the qualification method, environ-mental limit and function time, and the qualifying summary or report. The following systems are included.
MPL No. Title D21 Main Steam System B33 Reactor Recirculation System Cll Rod Control System C41 Standby Liquid Control System C51 Neutron Monitoring System 4 C61 Remote Shutdown System
- C71 Reactor Trip System D17 Process Radiation Monitoring System D23 Containment Atmosphere Monitoring System E12 Residual Heat Removal System E21 Low-Pressure Core Spray System E22 High-Pressure Core Spray System E31 Leak Detection System E32 MS Positive Leakage Control System E51 Reactor Core Isolation Cooling System Fil/13/15/16/42. Fuel Handling Equipment and Accessories l
G41 Fuel Pool Cooling and Cleanup System H13 Control Room Panels P38 Standby Gas Treatment System P45 Control Building Chilled Water System
. P53 Pneumatic Supply System P60 Water Positive Seal Isolation
() Leakage Control System 3.11-105 _
.,.,-f ,4-- _ . - _ ., - - - . . ., - . _ , , - ~ ,
, , , - , - , - . . , , - ,-y , - - --
GESSAR II ^
238 NUCLEAR ISLAND Rev. O Table 3.11-9 lh SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
MPL No. Title P61 Air Positive Seal Isolation Leakage Control System R43 Diesel Emergency Power T41 Reactor Building HVAC T49 Flammability Control Systen X63 Fuel Building HVAC X73 Auxiliary Building HVAC X93 Control Building HVAC XA3 Diesel Generator Building HVAC NOTES Qualification Method 1.
T = Testing A = Analysis T/A = Testing and Analysis f
= See Subsection 3.11.2.2.5.4.
Applies to equipment and supports
- 2. Environmental Limit Event Code Environmental Limit (EL)
A Abnormal B Small high energy (HE) pipe break in drywell and/or containment (SLB)
C Large HE pipe break in drywell D HE pipe break in adjacent zone E HE pipe break inside zone F Fuel handling accident O
3.11-106 _
.GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O Table 3.11-9 SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
(Continued)
Event Code Environmental Limit (EL) .
G Event B, C, or D*
II Event C or F*
I Event B or C*
J Event C, D, or E*
K Event D or E*
L Event C or D*
M Event not applicable, limiting normal conditions
- 3. Function Time FT Code Function Time (FT)
A 0 to 45 sec B 0 to 10 min C 0 to 1 hr D 0 to 6 hrs E O to 12 hrs F 0 to 24 hrs G 0 to 2 days H 0 to 30 days I O to 100 days
! J Always available (steady-state operation)
- 4. Qualifying Summary or Report Tables 3.11-9 (MPL-X) indicate a qualification summary table in this SAR. Other designations refer to a qualification report provided by the manufacturer.
- Satisfies any one of these conditions at any one time i (:)
m
' 3.11-107 _
l
Table 3. ll-9 (B21)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATIONS
SUMMARY
MAIN STEN 1 SYSTEM Qualifying Qualification Environmental Function Environmental Sunnary or Item No. Name Method 1 Limit 2 Time 3 Zone
- Report 4 B21 A001A-D Air accumulator B21 A002A-D Air accumulator B21 A003 Air accumulator (8)
B21 A004 Air accumulator (11)
N B21 D302 Condensing chamber w co B21 F022 Isolation valve ,,
w B21 F028 Isolation Valve b t* CD
+
H B21 F041 Valve, safety / relief yy b b21 F047 Valve, safety / relief Applicant to supply.
sM o CD H M B21 F051 Valve, safety / relief b
2 B21 G001 Main steam piping O B21 G002 Main stm pipe suspension B21 G003 Main steam P.W. restrnt B21 G005 Relief valva piping B21 G006 Main stm pipe suppressors B21 G009 Main steam guide / restraint B21 G024 Disch Quen X type ped mtd N
%M C0 >
<4
- O
- See Table 3.11-1 for identification of environmental zones. vi -J o
g -_ I 3.37 O O O
O O O
, Table 3.11-9(B21)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND EUVIRONMENTAL QUALIFICATIONS SG1 MARY -
MAIN STEAM SYSTEM (Continued)
Qualifying Qualification Environmental Function Environmental Summary or
. Item No. Name Methodl Limit 2 Time 3 Zone
B21 GG013 SRV line quencher B21 GG014 SRV line quencher B21 GG015 SRV line quencher U co B21 GG016 SRV line quencher 2:
7 B21 GG017 SRV line quencher @Q e us H tg tn H B21 GG018 SRV line quencher y>
t Applicant to supply. MM '
I
.$ B21 GG019 SRV line quencher B21 GG020 SRV line quencher 5U B21 GG021 SRV line quencher O ;
B21 GG022 SRV line quencher B21 GG023 SRV line quencher
, B21 GG024 SRV line quencher B21 GG025 SRV line quencher B21 GG026 SRV line quencher B21 GG027 SRV line quencher B21 GG028 SRV line quencher i
' B21 GG029 SRV line quencher w
- o w m>
i q
- See Table 3.11-1 fcr identification of environmental zones. I$O I
Un 4 i 1 3.37-
Table 3. ll-9 (B21)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATIONS
SUMMARY
MAIN STEAM SYSTEM (Continued)
Qualifying Qualification Environmental Function Environmental Summary or Item No. Name Method 1 Limit 2 Time3 Zone Report4 B21 K613 Power supply B21 K616 Millivolt to current conv B21 K617 Millivolt to current cony B21 K618 Millivolt to current cony B21 N004 Thermowell supp w/ element [
cn B21 N005 Flow element components u B21 NO27 Level trans @@
H t~ v1 B21 NO32 Diff press trans M u)
H >>
h B21 N040 Thermowell supp w/ element Applicant to supply. WW H mm O B21 N044 Level tranS CD m B21 N058 Press trans b g
B21 N061 Thermowell supp w/ element B21 N067 Press trans B21 N068 Press trans B21 N073 Level trans B21 N075 Press trans B21 N076 Press trans B21 N078 Press trans y
B21 N080 Level trans WM
- See Table 3.11-1 for identification of environmental zones. O U1 4 t I 3.37 O O O
. . ~ . . . ._ . - , -. - - . -. . - - .
.O O O-1 Table 3. ll-9 (B21) 1 SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATIONS
SUMMARY
MAIN STEAM SYSTEM (Continued)
Qualifying 5 Qualification Environmental Function Environmental Summary or Item No. Name Method 1 Limit 2 Time 3 Zone
- Report 4 B21 N081 Level trans B21 N091 Level trans B21 N094 Press trans B21 N095 Level trans N
B21 N097 Press trans co 4
3 B21 N099 Level trans 2 i
w B21 N658 Press indic sw gg H B21 N667 Press indic sw $$.
i APP l.icant to supply. mz
, e B21 N668 Press indic sw w
i H B21 N669 Press sw $U 7
- B21 N670 Press sw a
B21 N673 Level indic sw B21 N674 Level sw
- B21 N675 Press indic sw B21 N676 Press indic sw i
B21 N678 Press indic sw B21 N679 Press sw B21 N680 Level indic sw B21 N681 Level indic sw ,N (D >
- See Table 3.11-1 for identification of environmental zones. 4$o U1 4 4
0 t 3 l 3.37 a
Taole 3. ll-9 (B21)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATIONS
SUMMARY
MAIN STEAM SYSTEM (Continued)
Qualifying Qualification Environmental Function Environmental Summary or Name Methodl Limit 2 Time 3 Zone
- Report 4 Item No.
B21 N682 Level switch B21 N683 Level switch B21 N691 Level indic sw B21 N692 Level sw N
B21 N693 Level sw w co B21 N69'4 Press indic sw Applicant to supply. =
w c0
. B21 N695 Level indic sw o tu e hm y B21 N697 Press indic sw >>
WW e
e B21 N698 Press sw w $[
B21 N699 Level indic sw B21 R004 Press indicator O
B21 R005 Diff press indicator B21 R009 Diff press indic B21 R623 Level / press re' order
- See Tabic 3.11-1 for identification of environmental zones.
w WN O>
<4
- O O
U1 4 a
3.37 G G e
O O O Table 3.11-9 (B33)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
1 REACTOR RECIRCULATION SYSTEM Qualifying
! Qualification Environmental Function Environmental Summary or Item No. Name Method Limit Time Zone
- Report
}
B33 C001 Pump & motor 4
B33 D014 Sample probe I B33 F023 Gate valve-motor operated B33 F060 Flow control valve i
Gate valve-motor operated N B33 F067 w B33 G001 Recirc loop piping 2
W B33 GM2 Recirculation loop suspen CO
, O to l [ B33 G003 Recirc loop P. W. restr ($
8 B33 G006 Recirc loop suppressors Applicant to supply. WW
[
B33 N0ll Flow transmitter $[
B33 N014 Flow trans U
- I B33 NO21 Temperature element i
j B33 NO22 Temperature element B33 NO23 Temperature element B33 NO24
+ Flow trans
, B33 NO28 Tmp ele (dbl ele prec rtd) i B33 N029 Tmp Ele (dbl ele prec rtd)
B33 NO30 Tmp ele (dbl ele prec rdt) 1 g
- B33 NO37 Flow transmitter B33 NO38 Flow transmitter kN
<w
.. o I B33 N040 Pressure transmitter o W4 l *See Table 3.11-1 for identification of environmental zones. ,
4 r g
I 3.37 i
Table 3. ll-9 (Cll)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
ROD CONTROL SYSTEM Qualifying Qualification Environmental Function Environmental Summary or Item No. Name Methodl Limit 2 Time 3 Zone
- Report 4 C11 D001 liydraulic control unit Cll F009 Sol val >es for inst air C11 F010 Valve, globe, air operate Cll F0ll Valve, globe, air operate Cll N012 Pressure switch Applicant to supply. .
Cll N017 Leve' transmitter Z
. Cll N054 Pressure transmitter $@
t* in w to tn H
i Cll N601 Level switch . >>
WW
[ Cll N602 Level switch mm b tn H Cll N654 Pressure indicator switch b
g Cll N655 Pressure switch O
- See Table 3.11-1 for identification of environmental zones.
w
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- O O
(P 4 8
L-_-----
3.37 8 9 9
O O O
)
Table 3. ll-9 (C41)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
STANDBY LIQUID CONTROL SYSTEM Qualifying Qualification Environmental Function Environmental. Summary or Item No. Name Methodl Limit 2 Time 3 Zone
- Report 4 C41 A001 Storage tank C41 C001 Pump, standby liquid cont C41 C002s Motor standby lig contr C41 D003 Heater, mixin9 Applicant to supply.
C41 F004 Stby 1 q cont valve w m
C41 N003 Temperature switch w C41 N004 Transmitter, gage press $$
e tn
[ C41 N006 Control, temperature My b C41'R003 Indicator, pressure HH H
U Cn H h
- See Table 3.11-1 for identification of environmental zones. z o
l M
Al M 4
i -
8 wa B
3.37-
Table 3.11-9 (C51)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
NEUTRON MONITORING SYSTEM Qualifying Qualification Environmental Function Environmental Summary or Item No. Name Method 1 Limit 2 Timc3 Zone
- Report 4 C51 K001 Pulse preamplifier C51 K002 Voltage preamplifier C51 K600 Source range monitor C51 K601 Intermediate range monit N
C51 K605 Pwr range neut monit inst w C51 N001 Detector Applicant to supply.
z CO w C51 N002 Detector O trj H C51 N0ll Pwr range detector E$
i Oc lc r C51 N012 Pwr range detector C51 N013 Pwr range detector $gH C51 N014 Pwr range detector e
- See Table 3.11-1 for identification of environmental zones.
u
- c M C>
<4
. o U1 4 3.37 O O O
O O O a
Table 3.11-9 (C61)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
REMOTE SHUTDOWN SYSTEM 4 Qualifying Qualification Environmental Function Environmental Summary or Item No. Name Methodl Limit 2 Time 3 Zone
- Report C61 K001 Sq root converter C61 K002 DC-AC inverter C61 K005 Power supply C61 K010 Power supply
- Applicant to supply. "
C61 N001 Flow transmitter w oo C61 N006 Press transmitter w C61 N0lO Level indic trans $$
C61 P001 Remote shutdown vb Ny h ##
b:
w' C61 R001 Flow indic contr mH M Cf3 H
- *See Table 3.11-1 for identification of environmental zones.
h
.z i
O 4
4 1
e 1 ;
3 l
m
-%M m>
4 4
<w
.o i
1 O
, tn -a
+
l B 3.37 1
Table 3.11-9 (C71)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
REACTOR TRIP SYSTEM Qualifying Qualification Environmental Function Environmental Summary or Item No. Name Methodl Limit 2 Time 3 Zone
- Report 4 C71 N005 Pressure switch C71 N006 Turb stop vlv position sw C71 N050 Drywell high pr xmtr C71 N052 Turbine 1st stage pr xmtr Applicant to supply.
C71 P001 NSPS power distr encl N C71 P0ll NSPS power distr encl Z
C71 S001 Inv & static byp sw CO w
C71 S002 Static bypass switch assy $$
- O ,o
[
H mm O Cl) m
- See Table 3.11-1 for identification of environmental zones. b Z
O M
%M o>
<w
- O O
U1 4 0 8 3.37 O O O t -
1"% (3 n
( _.) (.) k Table 3.11-9 (D17)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
PROCESS RADIATION MONITORING SYSTEM Qualifying Qualification Environmental Function Environmental Summary or Item No. Name Methodl Limit 2 Time 3 Zone
- Report 4 D17 K609 Indicator & trip unit D17 K610 Log rad mon D17 K611 Indicator & trip unit Detector, insulated Applicant to supply.
D17 N003 D17 N009 Sensor & converter U co D17 N010 Isokinetic probe-bldg von =
cO W D17 NOll Sensor & converter OM
- t* Cn H M to H >>
I W :c H *See Table 3.11-1 for identification of environmental zones.
w tn H z
O N
- 0
, N o>
<: a
- O O
U1 4 i
I 3.37
Table 3. ll-9 (D2 3)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
CONTAINMENT ATMOSPHERE MONITORING SYSTEM Qualifying Qualification Environmental Function Environmental Summary or Item No. Name Method 1 Limit 2 Time 3 Zone
- Report 4 D23 K601 Log rad mon D23 K603 Pri contain/ hydrogen anal.
'123 N002 C::Unt hydro mon pnl D23 N003 Detector, insulated Applicant to supply.
D23 P600 Cams Ch A VB "
co D23 P601 Cams Ch B VB w
D23 R601 Recorder (type 521) $$
t< tn
[ D23 R603 Recorder (2 pen) gy 4
N lc :c O
ss Cn H
- See Table 3.11-1 for identification of environmental zones. g Z
O N
WN O :>
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- O O
U1 4 l l 3.37 O O O
O O O-Table 3. ll-9 (E12) s SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
RESIDUAL HEAT REMOVAL SYSTEM Qualifying Qualification Envircnmental Function Environ:cental Summary or Item No. Name Methodl Limit 2 Time 3 Zone
- Report 4 ,
E12 C002 RHR, pump & motor E12 C003 RHR, line fill pump E12 C009 Valve, motor-operated E12 C010 Valve manual,. globe E12 C001 RHR shtdn suction piping g E12 C006 RHR shutdn suct spnsn cn 2
E12 C008 RHR shutdown W
&, E12 N001 Conductivity element r tn tu rn Y E12 N002 Temp element >>
e Applicant to supply. M%
" E12 N003 Temp element HH g in H E12 N004 Temp element y Z
E12 N005 Temp element o E12 N007 Flow transmitter LI' N008 Level tranmaitter E12 N012 Flow orifice assy,ECCS E12 N013 Flow transmitter E12 N014 Flow orifice assy,ECCS [
E12 N015 Flow transmitter E12 NO26 Press transmitter E12 NO27 Temp element $$
< -J
. o
- See Table 3.11-1 for identification of environmental zones. o i
I' 3.37 i
Table 3. ll- 9 (E12 )
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
RESIDUAL HEAT REMOVAL SYSTEM (Co:'.tinued )
Qualifying Qualification Environmental Function Environmental Summary or Item No. Name Methodl Limit 2 Time 3 Zone
- Report 4 E12 NO28 Press transmitter E12 N050 Press transmitter E12 N051 Press transmitter E12 N052 Flow transmitter E12 N053 Press transmitter w co E12 N055 Press transmitter 2 w
E12 N056 Press transmitter $@
M U2 H tu r.n E12 N057 Press transmitter E12 N058 Press transmitter Applicant to supply. $$
N HH M tn H E12 N062 Press transmitter b
E12 N652 Flow ind switch 2 O
E12 N653 Press ind switch E12 N654 Press switch E12 N655 Press ind switch E12 N656 Press ind switch E12 N662 Press ind switch E12 R002 Press indicator E12 R008 Pressure indicator E12 R602 Flow indicator gw o>
E12 R603 Flow indicator <g O
m4
- See Table 3.11-1 for identification of environmental zones.
1 3.37 9 9 9
O O O Table 3.11-9 (E21)
SAFETY-RELATED EQUIPMENT IDEliTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
LOW PRESSURE CORE SPRAY SYSTEM Qualifying Qualification Environmental Function Environmental Summary or Item No. Name Methodl ' Limit 2 Time 3 Zone
E21 N002 Flow orifice assy, IPCS E21 N003 Flow transmitter E21 N049 Pressure transmitter y co i E21 N050 Pressure transmitter Z
=
w E21 N051 Flow transmitter @@
p cn E21 N052 Pressure transmitter to cn a
e -
E21 N053 ##
Y Pressure transmitter Applicant to Supply.
H HH N E21 N054 Pressure transmitter tn H E21 N649 Press ind switch %
g E21 N650 Press ind switch
- E21 N651 Flow ind switch E21 N652 Press ind switch E21 N653 Press ind switch E21 N654 Press ind switch E21 N655 Press switch 2
E21 N001 Pressure indicator w
E21 N002 Pressure indicator Ww E21 N600 Flow indicator $byo 4
- See Table 3.11-1 for identification of environmental zones, o .
- us .a !
1, i A 3.37-i l
Table 3.11-9 (E 2 2 )
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
HIGH PRESSURE CORE SPRAY SYSTEM Qualifying Qualification Environmental Function Envircnmental Summary or Item No. Name Methodl Limit 2 Timc3 Cone
- Report 4 E22 C001 HPCS pump E22 C003 liPCS line fill pump E22 C001 Motors, elec, vert mtd E22 C001 Elec motors (vert mounted)
E22 F001 Valve, motor operated [
co E22 F004 Valve, motor operated :2:
g E22 F010 Valve, motor operated QQ
'e E22 F0ll Valve, motor operated h$
E22 F012 Valve, motor operated Applicant to supply. $$
$ E22 F015 Valve, motor operated $[
E22 F023 Valve, motor operated t2 E22 N005 Flow Transmitter E22 N007 Flow orifice assy, HPCS E22 N008 Position transmitter E22 N010 Position transmitter E22 N050 Press trans E22 N051 Press trans E22 N052 Press trans E22 N054 Level trans yy E22 N055 Level trans k3
- O
- See Table 3.11-1 for identification of environmental zones.
i 3.37 9 9 e
O O O Table 3. ll-9 (E22)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
HIGH PRESSURE CORE SPRAY SYSTEM (Continued)
Qualifying Qualification Environmental Function Environmental Summary or Method 1 Limit2 Time 3 Zone
- Report 4 Item No. Name E22 N056 Flow trans E22 N650 Press ind switch E22 N651 Press ind switch E22 N652 Press ind switch w
w E22 N653 Press switch co E22 N654 Level ind switch Z CO gg W E22 N655 Level ind switch APP licant to supply. E$
E22 N656 Flow ind switch N:8
'" E22 R001 Pressure indicator 5U E22 R002 Press indicator E22 R603 Flow indicator O E22 R604 Position indicator E22 R606 Position indicator E22 S001 Eng-gen, HPCS E22 S002 Mot cont ctr, HPCS E22 S003 Transformer, HPCS E22 S004 Swgr elec, metal encl E22 AA01 Diesel Fuel Oil day tank Div 3 :o w o>
< -a
. o
- See Table 3.11-1 for identification of environmental zones. o t.n -a I
i !
i 3.37 1
Table 3. ll-9 (E31)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
LEAK DETECTION SYSTEM Qualifying Function Environmental Summary or Qualification Environmental Report 4 Item No. Name Method 1 Limit 2 Time 3 Zone
- E31 K600 Power supply E31 K602 Square root converter E31 K603 Square root converter E31 K604 Diff flow summer w w
E31 K605 Square root converter co E31 N001 Temperature element =
CO W am
, E31 N002 Temperature element t< tn Mm i
E31 N003 Temperature element .
Applicant to supply. yy
[ E31 N004 Temperature element MH MH E31 N005 Temperature element h
Z E31 N006 Temperature element O E31 N015 Temperature element E31 N017 Temperature element E31 N018 Temperature element E31 NO27 Temperature element E31 N028 Temperature element E31 NO29 Temperature element E31 NO30 Temperature element w
- c w E31 NO31 Temperature element o :P
<4
. o o
m
- See Table 3.11-1 for identification of environmental zones.
e i
3.37 8 9 9
._- _- . _ . --. . - . - - - .. -. __- . - - . _ _ . . . ~ . - - - - -
O O O
! Table 3. ll-9 (E31)
- SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
, LEAK DETECTION SYSTEM (Continued)
Qualifying I
Qualification Environmental Function ~. Environmental Summary or Limit 2 Report4 Item No. Name Method 1 Time 3- Zone
- _
E31 NO34 Temperature element -
E31 NO35 Temperature element E31 NO36 Temperature element -
E31 NO37 Temperature element M
E31 NO38 Temperature element ,
00 r E31 NO39 Temperature element ,
~
b) E31 N040 Temperature element @
I
[ E31 N041 Temperature element b$
i Applicant to supply. WM H E31 NO42 Temperature element
" HH i
" E31 N043 Temperature element _
to H E31 N044 Temperature element E-- b g
O l E31 N045 Temperature element ,
i E31 N046 Temperature element l
. E31 N047 Temperature element .
E31 N048 Temperature element E31 N049 Temperature element E31 N050 Temperature element )
E31 N051 Temperature element u
E31 N052 Ternperature element ww -
3 >.
.O O >
- See Table 3.11-1 for identification of environmental zones. L' 4 -
j h 3.37
Table 3. ll-9 (E31)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
LEAK DETECTION SYSTEM (Continued)
Qualifying Qualification Environmental Function Environmental Summary or Methodl Limit 2 Time 3 Zone
- Report 4 Item No. Name E31 N053 Temperature element E31 N054 Temperature element E31 N055 Temperature element E31 N056 Temperature element N
g E31 N057 Temperature element m E31 N075 Diff press trans g CO y E31 N076 Diff press trans O tu t< Cn E31 N077 Diff press trans Applicant to supply. $$
h y
E31 N083 Diff press xmtr HH MH m E31 N004 Diff press xmtr b
E31 N085 Press xmtr Z C
E31 N086 Diff press xmtr E31 N087 Diff press ymtr E31 N088 Diff press xmtr E31 N089 Diff press xmtr E31 N600 Diff temp switch E31 N602 Temp switch E31 N603 Diff temp switch N
- o N E31 N604 Temp switch (D >$
< -J
- o o
- See Table 3.11-1 for identification of environmental zones. vi -a f
3.37 O O e
O O O Tab]c 3. ll-9 (E 31)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
LEAK DETECTION SYSTEM (Continued)
Qualifying Qualification Environmental Function Environmental Summary or Method 1 Limit2 Time 3 Zone * . Report 4 Item No. Name E31 N605 Diff temp switch' E31 N608 Temp switch E31 N609 Dift flow switch E31 N610 Temp switch w
E31 N611 Diff temp switch w co E31 N612 Diff temp switch Z w E31 N613 Diff temp switch $@
[ E31 N614 Diff temp. switch Applicant to supply, py ben 5%
b E31 N615 Diff temp switch .HH N th H W E31 N616 Diff temp switch 2
E31'N617 Diff temp-switch O l E31 N618 Diff temp switch E31 N619 Diff temp swit'.:h E31 N620 Temp switch E31 N621 Temp switch E31 N622 Temp switch E31 N623 Temp switch E31 N624 Temp switch E31'N625 Temp switch ww o>
< -a
. o
- See Table 3.11-1 for identification of environmental zones. vi a i
i 3.37
Table 3. ll-9 (E31)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
LEAK DETECTION SYSTEM (Continued)
Qualifying Qualification Environmental Function Environmental Summary or Item No. Name Method 1 Limit 2 Time 3 Zone
- Report 4 E31 N626 Temp switch E31 N627 Temp switch E31 N680 Diff press ind switch E31 N681 Diff press ind switch N
E31 N683 Diff press ind sw w E31 N684 Diff press ind sw W Applicant to supply. $g E31 N685 Press ind sw gg b E31 N686 Diff press ind sw h$
- D :D
[ E31 Nti87 Diff press ind sw E31 NJ88 Diff press ind sw $[
E31 N039 Diff press ind sw U
E31 N6'.'O Diff press sw E31 N691 Diff press sw E31 R63* Time ind switch E31 Rf'7 Time ind switch
- See Ta'.,'le 3.11-1 for identification of environmental zones.
w
!O N O>
<w e O 3.37 O O O
- O O .O Table 3. ll-9 (C32)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
MS POSITIVE LEAKAGE CONTROL SYSTEMS Qualifying Environmental Function Environmental Summary or Item Jo. Name Qualifica{
Method ion Limit 2
Time 3 Zone
- Report 4 l E32 D001 Flow orifice assembly 3
E 3 2 D021 ~ Flow orifice assembly E32 K601 Power supply E3 2 K602 Summer N
E32 K603 Summer co j E32 K621 Power supply Z
w g
E32 K622 Summer Applicant to supply. $O y E32 K623 Summer E $'
2 g NM 1 w E3 2 M601 Timer
" HH E 3 2 M621 Timer mH E3 2 N001 Press transmitter E32 N002 Press transmitter E32 N003 Press transmitter E32.N004 Press transmitter E32 N005 Press transnitter I *See Table 3.11-1 for identification of environmental zones.
< xw 0>
<4
- O j O
- vi -a l
, i n 3.37
)
Table 3. ll-9 (E32)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICA'110N
SUMMARY
MS POSITIVE LEAKAGE CONTROL SYSTEMS (Continued)
Qualifying Qualification Environmental Function Environmental Summary or Item No. Name Method l Limit 2 Time 3 Zone
- Report 4 E32 N006 Flow orifice assembly E3 2 N007 Flow transmitter E3 2 NO21 Press transmitter E32 NO22 Press transmitter N
E3 2 N023 Press transmitter m
E 3 2 N024 Press transmitter :2:
w E32 NO25 Press transmitter CQ g
[ E3 2 N026 Flow orifice assembly b$
Applicant to supply. y$
E 3 2 NO27 Flow transmitter E3 2 N601 Press ind sw $N E3 2 N602 Press ind sw E3 2 N603 Diff press sw E3 2 N605 Press ind sw E3 2 N608 Flow ind sw E 3 2 N609 Diff press sw E32 N610 Diff press sw E 3 2 U621 Press ind sw E32 N622 Press ind sw N
E32 N623 Diff press sw xN O>
<4
- O O
- See Table 3.11-1 for identific ation of environmental zones. mw I I 3.37 9 9 e
- O O O 7
Table 3.11-9 (E32)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
4 MS POSITIVE LEAKAGE CONTROL SYSTEMS (Continued) 4 Qualifying Qualificettion Environmental Function Environmental Summary or Item No. Name Methodl Limit 2 Time 3 Zone
- Report 4 E32 N625 Press ind sw E32 N628 Flow ind sw
] E32 N629 Diff press sw
) E32 N630 Diff press sw . Applicant to supply.
N E3 2 R603 Press controller
{ *
} E 3 2 ' R607 Flow recorder
- Z E32 R623 Press controller ]Q i
w E3 2 R627 Flow recorder 'N$
H WW j 4
}
w *See Table 3.11-1 for identification of environmental zones. yH z
a O
i
\
i e
1 1
~
Q i
W tJ (D ,)*
<~1
- O i C
', U1 4
.I i I
,, a 3.37
Table 3. ll- 9 (E51)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONf1 ENTAL QUALIFICATION
SUMMARY
REACTOR CORE ISOLATION COOLING SYSTEM Qualifying Qualification Environmental Function Environmental Summary or Item No. Name !!ethodl Limit 2 Time 3 Zone
- Report4 E51 C001 RCIC pump E51 C002 Turb, st, RCIC drive E51 C003 RCIC line fill leump E51 D003 RCIC steam trap E51 F063 Valve, M-0 N E51 F076 Valve, M-O Z
E51 G001 RCIC/HX piping y QQ H E51 G006 RCIC/HX suspension N$
H
W%
Applicant to supply. gg A E51 G009 RCIC HX guide restraint CD H E51 K601 Sq root converter E51 N001 Flow orifice assy, RCIC E51 N003 Flow transmitter E51 N007 Press transmitter E51 N010 Level switch E51 NO35 Level transmitter E51 NO36 Level transmitter E51 NO37 Level switch E51 N050 Press transmitter N NN E51 N051 Plow transmitter j Wy
. 0 E51 N052 Press transmitter o tn 4
- See Table 3.11-1 for identification of environmental zones.
, t 3.37 O O O
4 O O O Table 3. ll-9 (E51) ,
SAFETY-RELATED LQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
I REACTOR CORE ISOLATION COOLING SYSTEM (Continued)'
Qualifying Qualification Environmental Function Environmental Summary or Item No. Name Methodl Limit 2 Time 3 Zone
- Report 4 E51 N053 Press transmitter
- E51 N055 Press transmitter E51 N056 Press transmitter E51 N635 Level ind switch N-
! E51 N636 Level ind switch w E51 N650 Press ind switch
=
w E51 N651 Flow ind switch CQ'
[ E51 N652 Press ind switch h$
i .
Appl 1 Cant to supply. px 1 H E51 N653 Press ind switch
.w E51 N654 Press switch $U E51 N655 Press ind switch.
E51 N656 Press ind switch E51 N659 Flow switch j E51 R001 Press indicator ;
I
% E51 R002 Press indicator i
E51 R003 Press indicator E51 R004 Press indicator E51 R005 Thermometer
' w E51 R600 Flow contr xw o>
E51 R606 Flow indicator [y ;
o-0 L14 4 l *See Table 3.11-1 for identification of environmental zones. ;
I
- 3.37
Table 3.11-9 (Fil/13/15/16/42)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
FUEL HANDLING EQUIPMENT AND ACCESSORIES Qual.tfying Qualification Environmental Function Environmental Summary or Item No. Name Methodl Limit 2 Time 3 Zone
- Report 4 Fil E001 Fuel prep machine Fil E002 New fuel inspection stand Fil E0ll General purpose grapple Fll E012 Jib crane N
Fll E014 Fuel handling platform w co Fll E017 Fue? handling platform W F13 5005 Head holding pedestal $$
[ F13 E008 Dryer sep strongback b$
i Applicant to supply. %%
F13 E009 Head strongback carcusel
[
- r$ u F14 E002 Control rod grapple h
g F15 E003 Refueling platform equip C
F15 E005 Auxiliary platform F15 E006 Refueling platform, RH F16 E006 In-vessel rack F16 E009 Defect fuel storage cont F16 E026 Module assy 13x17 type V F16 E030 Equipment storage rack F42 D001 Inclined fuel xfr tube w
F42 G001 Bellows Ww (D >
<: 4
. o o
- 4
- See Table 3.11-1 for identification of environmental zones.
E 1
3.37 O O O
e (3 D r s.J d V)
Table 3.11-9 (G41)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION SU'4 MARY -
FUEL POOL COOLING AND CLEANUP SYSTEM Qualifying Qualification Environmental Function Environmental Summary or Item No. Name Method 1 Limit 2 Time 3 Zone
- Report4 G41 A001 FPCCU drain tank G41 B001 Heat exchanger, fuel pool G41 C001 Pump, fuel pool G41 K600 Power supply M
G41 N003 Level switch cc G41 N004 Level switch 2
w G41 N005 Temperature element $@'
t< (n H G41 N005 Pressure switch to tn H >>
N*
h W
G41 N007 Pressure switch Applicant to supply.
HH 4 G41 N008 Temperature element tn H G41 N009 Flow orifice assembly h g
G41 N010 Flow orifice assembly G41 N011 Transmitter, diff press G41 N012 Transmitter, diff press G41 N013 Temperature element G41 N014 Temperature element G41 N015 Level switch G41 N016 Conductivity cell u
G41 N018 Transmitter, gage press xw (D >
G41 N019 Transmitter, gage press fy o
U1 -J
- See Table 3.11-1 for identification of environmental zones.
t E 3.37
Table 3. ll- i) (G 41)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
FUEL POOL COOLING AND CLEANUP SYSTEM (Continued)
Qualifying Qualification Environmental Function Environmental Summary or Item No. Name Method 1 Limit 2 Time 3 Zene* Report4 G41 NO20 Temperature element G41 NO23 Temperature element G41 N024 Transmitter, gage press G41 N026 Temp transmitter G41 NO30 Level switch y co G41 NO31 Level switch g
G41 NO32 Level switch CO O trj G41 N600 Level switch Applicant to supply, b y$
i :D lC H
W G41 N601 Level switch HH G41 N602 Temp switch mH G41 N604 Temp switch b Z
c G41 N606 Time ind switch G41 N607 Time ind switch G41 P001 Pump area panel G41 P002 Fuel pool pump area A VB G41 P003 Fuel pool pump area B VB G41 R001 Pressure indicator G41 R002 Pressure indicator G41 R005 Temp recorder :oU C>
< -4
- O o
- See Table 3.11-1 for identification of environmental zones. * 'd i e 3.37 O O O
\
Table 3. ll-9 (G41)
I SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
FUEL POOL COOLING AND CLEANUP SYSTEM (Continued) 1 Qualifying Qualification Environmental Function Environmental Summary or
! Item No. Name Method 1 Limit 2 Time 3 Zone
- Report 4 J G41 R601 Level indicater 4
! G41 R603 Temp indicator 1
i G41 R605 Temp indicator i Applicant to supply
- G33 N0ll Flow orifice assembly F3m element, (Venturi) M I G33 NO35 G33 N040 Flow orifice assembly i 2 w
G33 N043 Flow limiter (Venturi) QQ
- w rM u) 03 j W
{ 4 *See Table 3.11-1 for identification of environmental zones. $$
a --
C/3 -
2 U
i i
N
! MN (D >
' <: w
- O O ,
i U1 4 i t i l 3.37
Table 3. ll- 9 (H13)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
CONTROL ROOM PANELS Qualifying Qualification Environmental Functicn Environmental Summary or Item No. Name Methodl Limit 2 Time 3 Zone
- Report 4 H13 H200 Fire stops H13 P601 Reactor core cooling BB H13 P638 Cntmt atm mon instr pn1 A H13 P639 Cntmt atm mon instr pnl B N
H13 P642 Div 2 leak detection VB w cn H13 P651 Rod action cont cab.
Z w H13 P652 Rod action cont cab. $@
t< zn H13 P654 Div 2 MSIV lkg cont cab Applicant to supply. gg WW h
b H13 P655 Div i MSIV Ikg cont cab HH O Ill3 P661 NSPS div 1 cabinet Cn H H13 P662 NSPS div 2 cabinet b
g H13 P663 NSPS div 3 cabinet H13 P664 NSPS div 4 cabinet H13 P669 Neutron / process rad div 1 H13 P670 Neutron / process rad div 2 H13 P671 Neutron / process rad div 3 H13 P672 Neutron / process rad div 4 H13 P678 Standby information VB H13 P680 Prin plant cont console y
<4
.O O
- See Table 3.11-1 for identification of environmental zones.
I i 3.37 O O O
O O O Table 3. ll-9 (H13)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
CONTROL ROOM PANELS (Continued)
Qualifying Qualification Environmental Function Environmental Summary or Item No. Name Methodl Limit 2 Time 3 Zone
- Report 4 H13 P701 Term cab / core clo div 1/2 H13 P702 Tenn cab /BB div 3/N H13 P703 Term cab /reac cont H13 P704 Term cab / proc area rad H13 P706 Term cab /NSPS/NM div 1 U co H13 P707 Term cab /NSPS/NM div 2 W
, H13 P708 Term cab /NSPS/NM div 3 $$
w t* m trJ m w H13 P709 Term cab /NSPS/NM div 4 >>
i Applicant to supply. **
[ H13 P711- Tenn cab /crd cont / ann w HH H13 P714 Term cab /div 2 misc' MH h
H13 P715 Term cab /div 1 misc z O
H13 P719 Term cab /stdby info H13 P865 Div 1 flamm cont pnl H13 P866 Div 2 flamm cont pnl H13 U701 Flr module /reac core clg H13 U703 Flr module /reac cont H13 U704 Flr module / proc & area rad H13 U706 Flr module /NSPS div 1 U13 U707 Flr module /NSPS div 2 xU O>
< J
.O O
- See Table 3.11-1 for identification of environmental zones. vi -J t n 3.37
Table 3.11-9 (H13)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
CONTxOL ROOM PANELS (Continued)
Qualifying Qualification Environmental Function Environmental Summary or Item No. Name Methodl Limit 2 Time 3 Zone
- Report 4 H13 U70o Flr module /NSPS div 3 H13 U709 Flr module /NSPS div 4 H13 U711 Flr mc.dule/crd cont / ann H13 U714 Flr module /div 2 misc w
H13 U715 Flr module /div 1 misc w co H13 U717 Flr module /tip crd recirc g CO W H13 U719 Flr module /stndby info O tn
- LA in H13 U801 Lateral transition duct Applicant to supply.
H H13 U802 Longitudinal trans duct A HH WH H13 U803 Long tran duct, non-std h
Z H13 U804 Transition duct, remote t3 H22 POOL LPCS local panel H22 P002 Reac wtr clnup lcl pnl H22 P004 R V lvl & press lcl pnl A H22 P005 R V lvl & press lcl pnl C H22 P009 Jet pump local panel B H22 P010 Jet pump local panel A H22 P0ll Stand liq cont Icl pnl u
H22 P015 Main steam flow lcl pnl A mw o>
< -4
- O
- See Table 3.11-1 for identification of environmental zones, vi8 Q ,
3.37 O O O
O O o Table 3.11-9 (H22)
SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL QUALIFICATION
SUMMARY
CONTROL ROOM PANELS Qualifying Environmental Environmental- Summary or Qualification Function Report 4 Item No. Name Methodl Limit 2 Time 3 Zone
H22 P'021 RIIR B local panel H22 P024 HPCS local panel u
Main steam flow 1cl pnl B w H22 P025 co H22 P026 RV lvl & press lcl pnl D Applicant to supply. g W () t9 H22 P027 RV lvl & press Icl pnl B t* m 3
)
H22 P030 SRM & IRM preamp encl A-D hy H22 P031 SRM & IRM preamp encl A-D HH MH H22 2032 SRM & IRM preamp encl A-D h
- 2l H22 P033 SRM & IRM preamp encl A-D C H22 PO41 Main st flow lcl pnl D H22 PO42 Main steam flow lc1 pnl C H22 P055 RHR C local panel
- See Table 3.11-1 for identification of environmental zones.
I w
38
<aO WM i
f 3.37 I
Table 3. ll-9 (P3 8)
STANDBY GAS TREATMENT SYSTEM Env Qualification Item No. Description MQ EL FT Zone
- Table or Report P38-CC-001A&B SGTS exhaust fan P38-CC-003A&B SGTb heat removal fan Applicant to supply.
P38-DD-001A&B Standby gas treatment and unit N
(A co Z
CO W OM t^ in M Cn Table 3.11-9 (P45) $$
CONTROL BUILDING CHILI ED WATER SYSTEM yH Env Qualification C g
Item No. Description MQ EL FT Zone
- Table or Report U P45-AA-001A&B Chilled water expansion tank P45-CC-001A&B Chilled water pump Applicant to supply.
P45-CC-002A&B Booster pump P45-ZZ-001A&B Water chiller unit 4
- See Table 3.11--1 for identification of environmental zones. w lc w O>
<w o
O U1 4 1 1 3.37 O O O
O O O Table 3.11-9 (P53)
PNEUMATIC SUPPLY SYSTEM Env Qualification Item No. Description MQ EL FT Zone
- Table or Report P53-AA-001A&B Air receiver P53-AA-002 Air bottles I.pplicant to supply.
P53-DD-005A&B Air filter Table 3. ll-9 (P60)
.. WATER POSITIVE SEAL ISOLATION LEAKAGE CONTROL SYSTEM CO W
, Env Qualification h$
Item No. Description MQ EL FT Zone
- Table or Report hy i NN
[ P60-AA01 WPS. water supply tank APP licant to supply. gg vi to s b
z O
Table 3.ll-9 (P61)
AIR POSITIVE SEAL ISOLATION LEAKAGE CONTROL SYSTEM-Env Qualification Item No. Description EL FT Zone
- Table or Report
_ MQ_
P61-AA001A Air receiver Applicant to supply.
P61-AA001B Air receiver M
MN
- see Table 3.11-1 for identificatio:: of environmental _ zones.
o U1 4 1 a 3.37
Table 3.ll-9(R43)
DIESEL EMERGENCY POWER I
Env Qualification Item No. Description MQ EL 2 FT 3 Zone
- Table or Report 4 R43-AA-001A&B Diesel fuel oil da'y tank, Div. 1& 2 R43-S-001A&B Diesel generator, Div. 1& 2 R43-5-001A&B-1 Lube oil keep warm pump R43-S-001A&B-10 Lube oil cooler R43-S-001A&B-ll Lube oil electric heater R43-S-001A&B-12 Lube oil pump - engine driven $$
co R43-S-001A&B-13 Jacket water pump - engine driven 2 R43-S-001AsB-14 Jacket water pump - standby $$
M U2 R43-S-001A&B-15 Start air aftcooler rt bank M U2 g
g l R43-S-001A&B-16 Flame arrestor, day tank ,
[ Applicant to supply, s4 r, R43-S-001A&B-17 Start air comp left bank U) H os R43-S-001A&B-18 Generator space heater hg R43-S-001A&B-19 Start Air Receiver left bank R43-S-001A&B-2 Jacket water circ. pump R43-s-001A&B-20 Start air dryer, left bank R43-S-001A&B-21 Start air aftcoole_, left bank R43-S-001A&B-3 Jacket water electric heater R43-S-001A&B-4 Starting air comp., right bank R43-S-001A&B-5 Fael oil pump - engine driven R43-S-301A&B-6 Fuel oil pump - motor driven w w k>
R43-S-001Aco-7 Start air receiver, right bank j >
q R43-S-001A&B-8 Start air dryer, right bank o us .a R43-5-001A&B-9 Jacket water cooler
- See Table 3.11-1 for identification or envir7nmental zones. B 3.37 O O O
O O O Table 3.11-9 (T41)
REACTOR BUILDING llVAC I ErW Qualification Item No. Description MQ EL FT Zone
- Table or Report I T41 CC004A&B Shield Annulus exh/ rec fan Applicant to supply.
T41 C008A&B H nixing blower 2
i l ,
w .
W oo w Table 3.11-9 (T49) $o n en SAFETY-RELATED EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL ta (n
! [
4 QUALIFICATION
SUMMARY
NN 2 A FLAfDiABILITY CONTROL SYSTEM
" HH (n H
' Qualifying {
Z Qualification Environmental runction Environmental
- Summary or U
Item No. Name Method l Limit (2) Time (3) Zone Report" i
f T49 Z001 Thermal recombiner '
T49 ZOOl Power supply panel Applicant to Supply.
i ;
j T49 Z001 Control panel T49 ZZ001A&B Hydrogen recombiner
- 3 N
WM -
(D :>
<: a
- See Table 3.11-1 for. identification of environmental zones.
g (n -a i . i 4 3.37 4 r
Table 3.11-9 (XA3)
DIESEL CENERATOR BUILDING HVAC ENV Qualification Item No. Description MQ EL FT Zone
- Table or Report XA3 AHUO7 D6 Div. 3 swgr rm vent unit XA3 A11UO8 D6 Div. 3 swgr rm vent unit XA3 CC001A DG room supply fan XA2 CC001B Dr: room supply fan XA3 CC002A DG room supply fan y XA3 CC002B DG room supply fan Z
XA3 CC003A DG room receive fan CO O to H
XA3 CC003B DG room receive fan [$
y XA3 CC004A DG room exhaust fan 7.pplicant to supply. W#
$ XA3 CC004E. LG room exhaust fan yHH m p XA3 CC005A DG room exhaust fan >
Z XA3 CC0052 DG room exhaust fan O XA3 CC006A DG room exhaust fan XA3 CC006B DG room exhaust fan XA3 CC007 DG room supply fan XA3 CC008 DG room supply fan XA3 CC009 DG room receive fan XA3 CC010 DG room exhaust fan XA3 CC0ll DG room exhaust fan g XA3 CC012 DG room exhaust fan $$
<4
. O O
- See Table 3.11-1 for identification of envirenmental zones. ""
e i 3.37 0 0 0
I l
i O O O t j Table 3.11-9 (XA3) t 1.
1 DIESEL GENERATOR BUILDING HVAC (Continued) 4 ENV Qualification :
- Item No. Description MQ EL FT Zone * - Table or Report 1 r XA3 CCOl7
- Battery room exhaust' fan .
l Applicant to supply.
i XA3 CCOl8 Battery room exhaust fan T
+
Y
- See Table 3.11-1 for identification of environmental zones. [
1 = i w
w &o '
'OM H e in H M in H
l >> t u WW W HH (n H ;
. i
, z
.O 5
I 4
! t j :
! W E0 -
1 t
(D >
4 =J j e O 4 O i U1 =J l
i I 3.37
Table 3.ll-9(X63)
FUEL BUILDING HVAC Env Qualification Item No. Description MQ EL FT Zone
- Table or Report XG3 ECU01 A&B SGTS room cooling unit X63 ECUO2 A&B FPCCU pump room cooling unit Applicant to. supply.
X63 ECUO3 A&B SA exh ran room cooling unit X63 ECU05 Seal Sy;. room cooling unit w
w co
- See Table 3.11-1 for identification of environmental zones. g CO O t1 tA tn tu (n
.u >>
H lU :0 HH Y Cn s
~
U1 (
P O
Z O
w
%w 0>
<4
. O O
U14 i I 3.37 O O ,
O
- - . . - - . . . . .-- - . . . . - ~ . . . _ . . . -.. ..-.. . .-- _ . . . _ . .. .. .
- O O O Table 3.ll-9(X73)
AUXILIARY BUILDING HVAC 4
Env . Qualification Description MQ EL FT Zone
- Table or Report
, Item No.
X73 ACUO2 A&B AB electrical area ac unit X73 ACUO3 AB self contained ac unit X73 ACUO4 AB self' contained ac unit X73 ACUO5 Remote shutdown panel room cooler X73 CC012 A&B AB battery room exh fan w
AB battery room exh fan Applicant to supply, w I X73 CC015 B co I X73 ECUO3 LPCS pump room cooling unit .g
" CO RHRA pump room cooling unit nm j 'g X73 ECUO4 VM w MM ,
i
- i X73 ECUO5 RCIC pump room cooling unit >> ;
. p # '
S y X73 ECUO6 RHRC pump room' cooling unit HH
! mH X73 ECUO7 RHRB pump room cooling unit
] h X73 ECUO8 HPCS pump room cooling unit z
, O X73 ECUO9 Seal System room cooling unit t
1
- See Table 3.11-1 for identification of environmental zones.
\ '
4 w ww O>
<4 r
- O o
(n -J i I
< 3.37 i
4 ,
a
Table 3.ll-9(X93)
CONTROL BUILDING HVAC Env Qualification Item No. Description MQ EL FT Zone
- Table or Report X93 ACU01 A&B Control Bldg ac unit X93 AHU01 A&B Chiller room ac unit X93 CC002 A&B Control room ret /exh fan X93 CC003 A&B ODA cleanup unit supply fan Applicant to supply.
X93 CC004 A&B Control equip room ret /exh fan N
Normal intake tornado damper W X93 FF052A X93 FF052B Alt. intake tornado damper z CO w X93 OAC01 A&B Outdoor air cleanup unit pQ w m in H >>
I WW m *See Table 3.11-1 for identification of environmental zones. yH h
=
N NN O>
<4
. o O
U1 4 i i 3.37 O O O
GESSAR II 22A7007 238' NUCLEAR ISLAND Rev. 5 O Table 3.11-10 ,
J i
SELECTED GE POSITIONS ON NUREG-0588 E
O GE PROPRIETARY - provided under separate cover j l
O 3.11-153 through 3.11-160
GESSAR II' 22A7007 238 NUCLEAR. ISLAND Rev. 5
( Table 3.11-11 EXAMPLE OF A TEST SEQUENCE FOR TYPE TESTING P
Protest Inspection Baseline Functional Test under Normal Conditions , ,
Operational Test under Extremes of Functional Performance Characteristics Aging to End-of-Qualified Life Conditions 4
Baseline Functional Test under Normal Conditions O Operational Test under Dynamic Conditions.
Basel,ine Functional Test under Normal Conditions Operational Test under Design Basis Event Conditions Operational Test under Post-Design Basis Event Conditions Baseline Functional Test under Normal Conditions Post-Test Inspection Test Report 3.11-161 -
1
. . 1
GESSAR II 22A7007 230 NUCLEAR ISLAND Rev. 5
[)
b CT4 SLCS AREAS CT 1 ABOVE REFUELING FLOOR REFUELING FLOOR DW4 CT-5 RWCU ROOMS CT4 SLCS AREAS DW4 RPV DW1
/
CT 3 HCU FLOOR AND 20 ft ABOVE t\
U r,- 1 -
9 lCOREI 9 L_ _ J lI !l j- - CT 2 ABOVE SUPPRESSION POOL
-( (APPROX 20 f t ABOVE) r DW-5
' r DW-3 O O n ,n I . I DW-1 OUTSIDE RPV SHIELDWALL, NOT AT CORE MIDPLANE DW2 OUTSIDE RPV SHIELDWALL, AT CORE MIDPLANE EXTENDING TO A RADIUS OF 24 ft FROM RPV CENTERLINE, TO A HEIGHT 22 ft ABOVE FUEL CORE MIDPLANE AND 16 ft BELOW FUEL CORE MIDPLANE AND INCLUDING MAJOR RPV SHIELDWALL PENETRATIONS DW3 AREA UNDER RPV INSIDE PEDFSTAL DW4 WITHIN DRYWELL HEAD DW-5 OUTSIDE RPV SKIRT DW4 ANNULU6 BETWEEN RPV AND RPV SH!ELDWALL ,
O v
Pigure 3.11-1. Drywell and Containment Environmental Zones t
3.11-163/3.11-164 _
GESSAR II 22A7007 238. NUCLEAR ISLAND Rev. 5 j(]) 3A.3 INPUT. MOTION AND DAMPING VALUES 3A.3.1 Input Motion khe time-history method was used_in performing.the seismic analy-sis of the foundation / structures / reactor complex by using the finite-element method. Earthquake motion in the form of accelera-tion time histories for all three components was used. Since the input earthquake motion specified in Regulatory Guide 1.60 is in the form of response spectra, artificial earthquake acceleration time histories must be created based on the given spectra.
Numerous schemes have been proposed for synthesizing an earthquake
-12 However, the method pro-time history from a given spectrum .
7 posed by Vanmarcke and Cornell , was adopted here due to its intrinsic capability of imposing statistical independence among.
the synthesized time-history acceleration components. This is consistent with that given in Reference 3.
Three earthquake acceleration time-history components were created-by using this method and identified as Hy, H2, and V. These are the same time histories used in Reference 3. H y and H 2 were the-two horizontal components mutu' ally perpendicular to each other and were based on the US NRC horizontal ground spectra in Figure 3.7-1.
Hy and H2.are shown in Figures 3A-9 and 3A-10, respectively. V.
was the vertical earthquake time-history, component and was based on the vertical ground spectra in Figure 3.7-2. It is shown in-Figure 3A-ll. Subsection 19.3.3.31 provides the velocity and dis- g placement time histories corresponding to the input acceleration- 9 a
time histories. Figures 3.7-1 and 3.7-2 are for maximum ground 4
~
accelerations of 0.39 in the horizontal direction for the SSE
-condition.
Two conditions are to be met for the synthesized acceleration time histories. First, the response spectra generated from the time history must be a reasonable fit to the appropriate response O .
3A.3-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3A.3.1 Input Motion (Continued) lll spectra. Secondly, the three components shall be sufficiently statistically independent. .
In order to verify the first condition, response spectra with damping values of 1%, 2%, 3%, 4%, 7%, and 10% for all three com-ponents of the earthquake time history were generated. These response spectra, shown in Figures 3.7-4 through 3.7-21, are compared with the required spectra in Regulatory Guide 1.60. The peak acceleration of the time histories of these comparison curves was normalized to a horizontal acceleration of 0.15g OBE. The
. closeness of the two spectra in all cases indicates that the synthesized earthquake time histories are acceptable.
To check the statistical independence of the three earthquake components, the coherence function was used.
The coherence functions for the three earthquake acceleration
~
time-history components H1, H2, nd V were generated. The coher-ence function for H y and H 2, H y and V, and H 2 and V are given in Figures 3.1, 3.2, and 3.3, respectively, in Appendix 3A of Refer-ence 3. All values within the frequency range of interest between 0 and 50 Hz were calculated at a frequency increment of 0.1 Hz.
The small values of these coherence functions indicate that the three components are sufficiently statistically independent.
Figures 3A-12 through 3A-14 show the response spectra for the three earthquake time-history components at 2% damping for 0.15 OBE and the associated NRC Regulatory Guide 1.60 spectra.
3A.3.2 Damping The damping values for each component in the Reactor Vessel Building structure system used in this appendix are in accordance with those specified in Regulatory Guide 1.61. These values for l
both the OBE and the SSE conditions are summarized in Table 3.7-1.
3A.3-2
-GESSAR II '
22A7007 238 NUCLEAR ISLAND Rev. 5
() 3A.5.2 Analysis Procedure (Continued)
This approach is iterative and' requires about three to five iterations to obtain strain-compatible solutions.
Within three to four iterations, the soil properties were found to converge to a strain-compatible solution.
Analysis to obtain vertical components of deconvoluted s
bedrock motion was carried out without iterations using the final. soil properties from the analysis of the corresponding horizontal component case (e.g.,
GE-75-A-H2 for GE-75-A-V) for both free-field and finite-element analyses.
Response spectra plots and acceleration time-history responses at various node points on the basemat of the structures were generated from the finite-element
()
analysis. In addition, response spectra were also generated in the free field at top and bottom of soil layers'and at elevation of the Reactor Building basemat.
Response spectra for all the cases, generated at the center of top of basemat, are shown in Figures 3A-23 and 3A-24. These response spectra plots were used to compare the response with various other cases. The spectra in Figures 3A-23 and 3A-24 are well within the basemat envelope spectra.3 Both translational and rotational time histories of acceleration response generated at the top of basemat of structures were used to analyze a more detailed model cf the structures to evaluate their response and generate response spectra a,t critical elevations for analysis of equipment.
The SSE carthquake is taken as two times the OBE earth-quake (0.3g versus 0.15g) and the Regulatory Guide 1.60 m m
{} is used for defining them. Hence, no separate SSI analysis was performed for the SSE ihput.
3A.5-9/3A.5-10
O O O 1200 OBE 0.15g D AMPlNG 0.02
- - .. (UPE R - 75)
. . . . . . . . . ( AV E R - 75)
- - - (LOWR - 75)
(FIXH - H2)
{ .i
(
l
- --- (vP03 - 75) )
- (VPOS - 75) .
l g _.-.- (VPHR - 75) [g j j 5 - (PE RP - 75) l w
- W 2 --== (UPER - 150) f m
) 1 '
t Eo y 3 i
- I OM i
2 9
600 -
Y
,h em m 4 4 3 ,5 I . I 1 $$
ww 1 s \-..
d '.
! { HH 400 -
j .
- { l I **
200 -
g
-- uwa f
/
l 0 - 1 I I l !lI I I i i iIIl ! I l 1 I i1I 2
i 10' ' 10 iO' iO g F REQUE NCY (Hil , h Figure 3A-23. Finit.e-Elenient Horizontal Response Spectra at Top of Basemat .
- o o
O4 I
l
oMmsW HH MM>W NE zcO >W HC9z O
WO<
0
! O 1
t
_ a m
- e s
- a B
s n f o
~
p
- o j , T
- t m a g\ a
\ 0 1
r N t 4 \ c e
Q* \ p S
\
k 1 \ t e 2 \ I )
z s
0 0 H n
\
(
o gG
\ ' Y p 5 C s 1 N N OPI e 5 3O E
EM / ' U R BA f Q .
OD I / E R l 3 F a c
i t
r e
V t
0 n 0
1 e
m e
' l E
e t
i n
_ i
)
5 ! F 7 -
ivR x EV !
4
_ i F
(
A
( 2
- A
- I 3
- e r
- u g
i I F
- - - - -0 1
0 O 0
0 8 m EIEg$- 8a< E 58 e 0 2
9 l
u>. ?M
_ _ = - . _ . _
m GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
~
O s
APPENDIX 3I ENVIRONMENTAL QUALIFICATION TESTING EFFORT ADMINISTRATIVE CONTROLS I
4 O .
l l
O 1
s GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
./'N
(.)
APPENDIX 3I CONTENTS Section Title Page 3I ENVIRONMENTAL QUALIFICATION TESTING EFFORT .
ADMINISTRATIVE CONTROLS 31.1 PRODUCT PERFORMANCE QUALIFICATION SPECIFICATION GUIDELINES 31.1-1 31.1.1 Scope 3I.1-1 3I.1.2 Documents 3I.1-1 3I.1.2.1 Applicable Documents 3I.1-1 31.1.2.2 Reference Documents 3I.1-1 31.1.3 Performance Specification 3I.1-2 3
3I.l.3.1 Product Description 3I.1-2 3I.1.3.2 Interface Definition 3I.1-2 3I.l.3.3 Mounting 3I.1-2 3I.1.3.4 Performance Requirements 31.1-2 I'l
%j 31.1.3.5 Physical Characteristics 3I.1-3 31.1.3.6 Maintenance Requirements 31.1-3 3I.1.3.7 Service Conditions 3I.1-3 3I.1.4 Qualification Specification 3I.1-3 3I.l.4.1 Test Sample Description 3I.1-4 3I.l.4.2 Quality Assurance 3I.1-4 3I.1.4.3 Set-Up Requirements 3I.1-4 31.1.4.4 Test Sequence 3I.1-4 3I.l.4.5 Baseline Tests 3I.1-4 3I.1.4.6 Abnormal Conditions 3I.1-5 31.1.4.7 Aging 3I.1-5 3I.l.4.8 Dynamic Test 3I.1-5 3I.1.4.9 Design Basis Event Test 3I.1-5 31.1.4.10 Disassembly and Inspection 3I.1-5 31.1.4.11 Acceptance criteria 3I.1-5 31.1.4.12 Test Report 3I.1-6 Test Specimen Deposition 3I.1-6
(} 3I.1.4.13 3I-i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O
CONTENTS (Continued)
Section Title Page 3I.2 PRODUCT ANALYSIS REPORT GUIDELINES 31.2-1 3I.2.1 Scope 31.2-1 31.2.2 Report Format 3I.2-1 3I.2.2.1 Cover Sheet 3I.2-1 31.2.2.2 Table of Contents 3I.2-1 31.2.2.3 Scope 3I.2-1 31.2.2.4 Product Description 3I.2-1 3I.2.2.5 Environment 3I.2-2 3I.2.2.6 Summary of Results 3I.2-2 31.2.2.7 Detailed Analyses 3I.2-2 31.2.2.7.1 Aging Analysis 3I.2-2
~
3I.2.2.7.2 Similarity / Traceability Analysis 3I.2-3 3I.2.2.7.3 Other Analyses 3I.2-4 3I.2.2.8 Acceptance Criteria 3I.2-4 3I.2.2.9 References 3I.2-4 3I.3 SIMILARITY / TRACEABILITY PROCEDURE 3I.3-1 31.3.1 Scope ,
3I.3-1 31.3.2 Establishment of Family of Products 31.3-1 3I.3.2.1 Manufacturer 3I.3-1 3I.3.2.2 Function 3I.3-1 3I.3.2.3 Quality Assurance Program 3I.3-1 3I.3.2.4 Functional Mechanism 31.3-2 3I.3.3 Similarity Analysis 3I.3-2 31.3.3.1 Service / Application and Configuration 3I.3-2 3I.3.3.2 Range / Rating 3I.3-2 3I.3,3.3 Interface 3I.3-3 3I.3.3.4 Stresses 3I.3-3 3I.3.3.5 Materials 3I.3-4 31.3.4 Summary 3I.3-5 0
GESSAR II 22A7007 .
238 NUCLEAR ISLAND Rev. 5
/"'s V
CONTENTS (Continued)
Section Title Page 3I.4 FORMAT A D CONTENT OF PRODUCT APPLICATION FUNCTIONAL REQUIREMENTS 31.4-1 3I.4.1 Purpose 31.4-1 31.4.2 Introduction 3I.4-1 3I.4.3 Format 31.4-1 31.5 PRETEST EVALUATION 3I.5-1 3I.5.1 Scope 31.5-1 31.5.2 Requirements 3I.5-1 3I.5.2.1 Test Conditions 31.5-1 3I.5.2.2 Functional Requirements 31.5-1 3I.5.2.3 Product Description 3I.5-2 31.S.2.4 Method of Evaluation 31.5-2 3I.S.2.4.1 Manufacturer's Data 3I.5-2 O)
(__ 31.5.2.l.2 Test Data 31.5-3 3I.S.2.4.3 Evaluation by Operating Experience 3I.5-4 31.5.2.4.4 Evaluation by Ccmponent 3I.5-4 31.5.2.4.5 Recommendations 3I.5-5 3I.5.3 Documentation 31.5-5 31.5.3.1 Records 3I.5-5 31.5.3.2 Evaluation Report 3I.5-6 31.5.4 Appendices 3I.5-6 3I.6 TEST PLAN AND PROCEDURES GUIDELINES 3I.6-1 3I.6.1 Scope 31.6-1 31.6.1.1 Purpose 3I.6-1 3I.6.1.2 Use 3I.6-1 3I.6.2 Document Control 3I.6-1 3I.6.3 Format and Content 3I.6-1 3I.6.3.1 Cover Sheet 3I.6-2 31.6.3.2 Table of Contents 3I.6-2 3I.6.3.3 Test Plan 3I.6-2 A 31.6-4
'q,) 3I.6.3.4 Test Procedure 3I-iii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O
4 CONTENTS (Continued)
Section Title Page
~
3I.7 TEST REPORT GUIDELINES 3I.7-1 31.7.1 Scope 3I.7-1 3I.7.2 Report Format 31.7-1 3I.7.2.1 Cover Sheet 3I.7-1 3I.7.2.2 Table of Contents 3I.7-1 3I.8 ENVIRONMENTAL QUALIFICATION REPORT GUIDELINES 31.8-1 O
.h O
GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 5 O
k./
APPENDIX 3I .
ENVIRONMENTAL QUALIFICATION TESTING EFFORT ADMINISTRATIVE CONTROLS a
This appendix provides examples of the administrative controls which GE implements on its environmental qualification testing effort. They are not considered specific commitments, rather, they are representative of the level of administrative controls which GE feels are needed to be available in order to adequately plan for, perform, and document the specific qualification efforts to which this report specifically will be applied. Each section of this appendix is a separate administrative control as listed below:
Section Administrative Control
- p. 31.1 Product Performance Qualification Specification i L/ Guidelines 3I.2 Product Analysis Report Guidelines 3I.3 Similarity / Traceability Procedure 3I.4 Format and Content of Product Application Functional Requirements 3I.5 Pretest Evaluation 3I.6 Test Plan and. Procedures Guidelines 3I.7 Test Report Guidelines
, 3I.8 Environmental Qualification Report Guidelines
- 3I-v/3I-vi
l GESSAR II 22A7007 238 NUCLEAR ISLAND 'Rev. 5 O
31.1 PRODUCT PERFORMANCE QUALIFICATION SPECIFICATION GUIDELINES 1
O PROPRIETARY INFORMATION.- provided under separate cover 1
t 0
3I.1-1 through 3I.1-6
GESSAR II 238 NUCLEAR ISLAND 5 s
O 3I.2 PRODUCT ANALYSIS REPORT GUIDELINES
.1 1
l PROPRIETARY INFORMATION - provided under separate cover O
3I.2-1 through 3I.2-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
- O 3I.3 SIMILARITY / TRACEABILITY PROCEDURE l
O raoea er^ar "roa"^r on - vroviaea eaaer eeeerete cover l
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O
3I.4 FORMAT AND CONTENT GUIDE OF PRODUCT APPLICATION FUNCTIONAL REQUIREMENTS s) PROPRIETARY INFORMATION - provided under separate cover I
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O 3I.4-1 through 3I.4-6
. . - . . - . - . - - - . - - . .~. -_ - - . . . . - - -
! GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 l
3I.5 PRETEST EVALUATION i
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O
l 3I.6 TEST PLAN AND PROCEDURES GUIDELINES PROPRIETARY INFORMATION - provided under separate cover o
3I.6-1 through 3I.6-6
- .- -- -- __. . _ _ _ . -_ _ . _ _ _ - . _ - _ _ _ . _ _ . _ __ \
. . . _ = - = . . . - . - - - - - .. _. - . . -. - .
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GESSAR II 22A7007 ,
238 NUCLEAR ISLAND Rei . 5 .
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GESSAR II '
22A7007 238 NUCLEAR ISLAND Rev. 5 O
3I.8 ENVIRONMENTAL QUALIFICATION REPORT GUIDELINES l
4 PROPRIETARY INFOR"ATION - provided under separate cover 1
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GESSAR II 22A7007'
. 238 NUCLEAR ISLAND Rev. 0
) 7.1.2.3.18 ' Control Building Chilled Water System (CB-CHILL)
Instrumentation and Controls A. Safety Design Basis ,
i l 1. General Functional Requirements The general functional requirements of the Control Building Chilled Water System instrumentation and controls shall be to function under,,all modes of plant operation and accidents as necessary to ensure a continuous supply of chilled water to the cooling coils of the Control Building air conditioning units and chiller room coolers and the Auxiliary Building ele'ctrical switchgear room air conditioning system.
- 2. Specific Regulatory Requirements O The specific regulatory requirements applicable to the instrumentation and controls of the CB-CHILL are given in Tabic 7.1-3.
B. Nonsaf'ety-Related Design Basis The CB-CHILL shall provide a continuous supply of chilled water to i the cooling coils of air conditioning systems which provide a controlled temperature environment and proper humidity to ensure i the comfort of the operating personnel and to provide a suitable atmosphere for the operation of control equipment.
1 4
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l 7.1-39 i
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 7.1.2.4 Safe Shutdown Systems - Instrumentation and Control ggg 7.1.2.4.1 Reactor Core Isolation Cooling (RCIC) System Instrumantation and Controls A. Safety Design Bases
- 1. General Functional Requirements The general functional requirements of the instrumenta-tion and controls of the system are to:
(a) initiate RCIC System operation as necessary to maintain sufficient coolant in the reactor vessel in case of an isolation with a loss of main feedwater flow; (b) provide for automatic and remote manual operation of the system; (c) satisfy Seismic Category I design requirements g (all components with Safety Function);
(d) provide assurance that the RCIC I&C shall operate when necessary (startup of the system shall not be dependent upon auxiliary AC power or plant air and external cooling water systems);
(e) provide assurance that the system shall operate when necessary by periodic testing that can be performed during plant operation.
(f) mitigate the consequences of a control rod drop accident. _
- 2. Specific Regulatory Requirements The specific regulatory requirements applicable to the control and instrumentation of this subsystem are given llh in Table 7.1-4.
7.1-40
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O 7.1.2.4.1 Reactor Core Isolation Cooling (RCIC) System Instrumentation and Controls (Continued)
FCIC'is considered an En'gineered Safety Feature System because it mitigates the' consequences of a control room rod drop accident. .
B. Nonsafety-Related Design Basis None 7.1.2.4.2 Standby Liquid Control System (SLCS) Instrumentation and Controls A. Safety Design Basis
- 1. General Functional Requirements The general functional requirements of this equipment O are to provide necessary control of the SLC equipment for shutting the reactor down from full power to cold shutdown and maintaining the reactor in a suberitical state at atmospheric temperature and pressure conditions by pumping sodium pentaborate, a neutron absorber, into the reactor.
- 2. Specific Regulatory Requirements The specific regulatory requirements applicable to this system are given in Table 7.1-4.
B. Nonsafety-Related Design Basis None O
7.1-41
GESSAR II 238 NUCLEAR ISLAND 22A7007 Rev. 0 7.1.2.4.3 RHR - Peactor Shutdown Cooling Mode (RHRS) -
Instriu'tentation and Controls lh A. Safety Design Bases
- 1. General Functional Requirements The general functional requirements of the RHRS are to provide monitoring and control as required to: *
(a) enable the system to reraove the residual heat (decay heat and sensible heat) from the reactor vessel during normal shutdown; (b) provide manual controls for the shutdown cooling system in the main control room and at the remote shutdown panel; and (c) Indicate performance of the shutdown cooling system by main control room instrumentation and instrumentation in the remote shutdcwn panel.
- 2. Specific Regulatory Requirements The specific regulatory requirements applicable to reac-tor shutdown cooling are given in Table 7.1-4.
B. Nonsafety-Related Design Basis The I&C shall provide monitoring and control to enable the RHRS to accomplish the following:
- 1. provide cooling for the reactor during the shutdown operation when the vessel pressure is below approxi-mately 135 psig; e
7.1-42
. - . . - - . - . . . . _ _ - . . - . - _ . _ - - . . . . _ . _ - - _ _ - ~ . . . _.. . . -
! O O O
- Table 7.1-1 DESIGN AND SUPPLY RESPONSIBILITY 1 i I Systems Designer Supplier Reactor Protection System ,
Reactor Protection System (RPS) GE GE I
j Engineered Safety Featured Systems ,a Emergency Core Cooling Spray (ECCS) GE GE
! High-Pressure Core Spray (HPCS) GE GE ,
, w y
Automatic Depressurization System (ADS) GE GE Low-Pressure Core Spray (LPCS) GE 'GE z CO 7' Low-Pressure Coolant Injection.,(LPCI) GE GE Oy 8 Containment and Reactor Vessel Isolation Control GE GE/U Em l 2$ System (CRVICS) WW
' HH l Main Steamline Positive Leakage Control System (MSPLCS) GE GE/U mH f Containment Spray Cooling (CS-RHR) GE GE
} Suppression Pool Makeup System (SPMU) GE U ,
Containment Combustible Gas Control System (CCGCS) GE U
! Standby Gas Treatment System (SGTS) GE U
- Shield Building Annulus Mixing GE U Secondary Containment Isolation Control System GE U l
- Containment Isolation Valve Leakage Control Systems u
l Air Positive Seal (APS) GE U
- yy Water Positive Seal (WPS) f
- GE U . o i o i
- W4 l --
'l i
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Table 7.1-1 s DESIGN AND SUPPLY RESPONSIBILITY (Continued) -
Systems Designer Supplier Standby Power System HPCS Diesel Generator System GE U Emergency Diesel Generator System GE U Diesel Generator Auxiliaries GE U Essential Service Water System (ESWS) GE U ESF Area Cooling System GE U w Pneumatic Supply System GE U m Main Control Room Heating, Ventilating, and Air Conditioning GE U $o
," system og w Y en i >>
w 'xx
- Systems Required for Safe Shutdown gg Reactor Core Isolation Cooling (RCIC) System GE GE U/GE U/GE f
Standby Liquid Control System (SLCS) $
RHRS/ Reactor Shutdown Cooling System GE GE Remote Shutdown System (RSS) GE U/GE Other Safety Systems Neutron Monitoring System (NMS)
Source Range Monitor (SRM)2 GE GE Inter:rediate Range Monitor (IRM) GE GE w xw Local Power Range Monitor (LPRM) GE GE >
uiO 9 O O
O O O Table 7.1-1 i ' DESIGN AND SUPPLY RESPONSIBILITY (Continued)
Systems Designer Supplier Average Power Range Monitor (APRM) GE GE Traversing Incore Probe (TIP) 2 GE GE Process Radiation Monitoring System (PRMS) GE GE Rod Pattern Control System (RPCS) GE GE High-Pressure / Low-Pressure Systems Interlock Function GE U Recirculation Pump Trip (RPT) System GE GE/U M Fuel Pool' Cooling and Cleanup System GE U
=
Drywell/ Containment Vacuum Relief System GE U gQ
'r, Containment and Reactor / Auxiliary / Fuel Building Ventilation GE U E$
3
$ and Pressure Control System y$
- 1 Containment Atmosphere Monitoring System GE/U U/GE ss
'j mH Suppression Pool Temperature Monitoring GE U $
z
, Reactor Vessel Instrumentation (partial) GE GE o
! Control Systems Not Required For Safety I
Reactor Vessel Instrumentation (partial) GE GE Rod Control and Information System Rod Movement Control- GE GE
< 4
,
- O W
I f f i
Table 7.1-1 DESIGN AND SUPPLY RESPONSIBILITY (Continued)
Systems Designer Supplier Pressure Regulator and Turbine Generator System U U Performance Monitoring System GE GE Reactor Water Cleanup (RWCU) System ,
GE GE Radwaste System Gaseous Radwaste System GE U Liquid Radwaste System GE U w GE U
- Solid Radwaste System Area Radiation Mon'itoring System ( ARMS) GE U 5o om 4
g Leak Detection System GE GE/U QQ E Containment Exhaust GE U $$
o Drywell Purge GE U y[
Suppression Pool Cleanup (SPCU) GE U U C Fire Protection System GE/U Breathing Air System GE U Drywell Chiller GE U Instrument Air j GE U Display Control System GE GE Refueling Interlock Function l GE GE N
NOTES %N
- 1. For mitigation of the rod drop accident only -
$3
- O The source range monitor and traversing incore probe are included in Neutron o
- 2. g Monitoring System discussion for completeness only; they are not safety subsystems.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 7.3-16 SECONDARY CONTAINMENT ISOLATION INSTRUMENTS AND RANGES Instrument Isolation Function Instrument Range Reactor vessel Level -160 to 60 inches Low water level Transmitter (System initiation)
Drywell high Pressure 0 to 5 psig Pressure Transmitter (System initiation)
O N_,)
N 4
v 7.3-417
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 Table 7.3-17 g AUXILIARY BUILDING ESF AREA COOLING SYSTEMS INSTRUMENTS AND RANGES Instrument ESF Function Instrument Range ECCS area Differential -1 to +0.5 differential pressure pressure inches water transmitter RWCU pump Temperature 40 to 180 0 F room high switch temperature RHR pump Temperature 40 to 160 0 F room high switch temperature LPCS pump Temperature 40 to 160 F room high switch temperature HPCS pump Temperature 40 to 1600F room high switch temperature Temperature 40 to 1600F O
RCIC pump room high switch temperature _
Electrical switchgear Temperature 0 to 150 F -
area high controller temperature Auxiliary building ECCS Radiation 0.10 to 100 mr/hr HVAC exhaust detector high radiation O
7.3-418
. ~ . - . .. .- - -- - . . - - . - .
GESSAR II 22A7007
- 238 NUCLEAR ISLAND Pov. 0 Rod Pattern Control' System (Continued)
-() 7.6.2.4.2.C Paragraph 4.16: Rod motion following.the recovery of the j rod motion permissive signal requires deliberate operator action.
Paragraph 4.17: A rd'd block may be manually initiated by the operator. Single failures cannot prevent manual.or automatic initiation since the system is fail-safe and a number of ways to ,
remove the rod motion permissive signa are available.
Paragraph 4.18: The design permits administrative control 4
of access to setpoint adjustments and testpoints.
Paragraph 4.19: The rod block is indicated and identified down to the division level.
1 Paragraph 4.20: The RC&IS provides the operator with accurate, complete, and timely information pertinent to the RPC
( )-
status.
\i Paragraph 4.21: The system is designed to. facilitate the recognition, location, replacement or repair of malfunctioning components or modules.
i
(
Paragraph 4.22: Portions of the RC&IS that are part of the
, essential system are identified distinctively as being in the protective system. Distinctions are made between redundant i divisions of the system.
,! Paragraph 4.21: The RPC is designed to facilitate the recog-j nition, location, replacement, and repair of malfunctioning i I
l 7.6-103 J
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 7.6.2.4.2.C Rod Pattern Control System (Continued) components and modules.' Failures can be readily traced to the rcodule or circuit card level with the aid of operator display and maintenance display indicators located throughout the system.
IEEE 323-1974 With exception of the individual rod position indication probes, the equipment is qualified, by test or analysis, in accor- J dance with IEEE 323-1974 to meet the environmental conditions of Tables 3.11.'3 and 3.11.4, except that the equipment function is that of a control system (i.e., does not operate on design basis event signals) and as such is not required to operate during or _
after a design basis event. The probes have been analyzed and verified to show that no credible failures can disable the correct execution of the essential rod block and pattern control functions.
This, combined with the " fail-safe" nature of the logic assures the overall safety integrity of these functions. _
On-line testability of the system and indication of bypassed or inoperable status of the system is provided (See Sub-section 7.6.1.4.D).
The system can be routinely tested in any rod configuration for that particular configuration, by observing the individual insert permissive and withdraw permissive signals. Any rod can oc selected and the correlation of the insert and withdraw per-missives can be observed and verified. By initiation of rod motion request (s) and proper action of the rod withdrawal, rod insertion, insert block, and withdraw block can be verified.
O 7.6-104
22A7007 GESSAR II 238 NUCLEAR ISLAND Rev. 5 A
V 7.6.2.4.2.C Rod Pattern Control System (Continued) a IEEE 344-1975 -
With exception of the individual rod position indication probes, the equipment is qualified, by test or analysis, in accordance with IEEE 344-1975 (see Section 3.10) . The probes are seismically qualified for the operational base earthquake (OBE) based on plant availability considerations and not safety con-siderations. The probes have been analyzed and verified to show that no credible failures can disable the correct execution of the
! essential rod block and. pattern control control functions. This combined with the " fail-safe" nature of the logic assures the overall safety integrity of these functions. _
O 1
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7.6-104a
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O - x
~ TIP CAllBRATION TUBE i
LPRM DETECTOR DETECTOR THIMiLE e4 3h F CF L BUNDLE CONTROL HOD BLADE
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GESSAR II 238 NUCLEAR ISLAND 22A7007 Rev. 5 O
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 0
LPRM STRING PROVIDING INPUT TO APRM CHANNELS A, C, E, AND G LOWER RIGHT LETTER = INPUT FOR APRM CHANNEL E UPPER RIGHT LETTER = INPUT FOR APRM CHANNEL G APRM TRIP SYSTEM A 0'
s3 0 YYO O YOO YYOOO 4 0 Y O O Y + 00
!!siiH H H H H H H O
[ YY OO O YYY O OU O O 000 00"
!!$iH H H H H H H irlH H H H H HiF oi "81H44H H H P O O 0O0O0000 00 00+0 0 00+0 0_tOOOO o3, m oi_ 0 0'0 0 I I I I I I I II i i 1 00 02 04'06 08 10 12 14 16 18 20 22 24 26 26 30 32 34 36 38 40 42 44 46 48 50 52 54 56 58 60 01 03 05 07 09 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 53 55 57 50
, Figure 7.7-13a. LPRM Channel Arrangement in the Core and APRM Channel Assignments 7.7-165
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 PPE 1 LEFT LE E = IPJPUT FOR A RM CHANNEL B LOWER LEFT LETTER = INPUT FOR APRM CHANNEL D ER F HT LETTER = l PUT FOR APRM CHANNEL H APRM TRIP SYSTEM B O
00000 000 60 ,
58 OOOOODOO Y1 4 00000 O O C+ 0 0 +0 0 4
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O d O O YO O O O O Y O Y O 02 0 0 0 E I I I I I I I I I I I I 00 02 04 06 08 10 12 14 16 18 20 22 24 26 28 30 32 34 36 38 40 42 44 46 48 50 52 54 56 58 60 01 03 05 07 09 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 53 55 67 59 Figure 7.7-13b. LPRM Channel Arrangement in the Core and APRM Channel Assignments 7.7-166
t 4
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! GESSAR II 22A7007
- 238 NUCLEAR ISLAND Rev. 5
! SECTION 16.1 I 4
1
- CONTENTS l l
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1 i Section Title Page i 16.1 DEFINITIONS 16.1-1 I
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CHAPTER 16 TECHNICAL SPECIFICATIONS ]
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GESSAR II 22A7007 238 13UCLEAR ISLAND Rev. 5 SECTION 16.2 CONTENTS Section Title Page 16.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 16.2-1 16.2.1 Safety Limits 16.2-1 16.2.2 Limiting Safety System Settings 16.2-3 ,
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4
- GESSAR II 22A7007 4
238 NUCLEAR ISLAND Rev. 5 i i
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GESSAR II 22A7007
- 238 NUCLEAR ISLAND Rev. 5 i
SECTION 16.B2 CONTENTS Sec' tion Title Page i
16.B2 BASES FOR SAFETY LIMITS AND LIMITING SAFETY 16.B2-1 ,
SYSTEM SETTINGS 16.B2.1 Safety Limits 16.B2-1 16.B2.2 -
Limiting Safety System Settings 16.B2-11 i
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4 GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O SECTION 16.3/16.4 CONTENTS
- Section Title Page 16.3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 16.3/4.0-1 16.3/4.0 Applicability 16.3/4.0-1 16.3/4.1 Reactivity Control Systems 16.3/4.1-1 16.3/4.2 Power Distribution Limits 16.3/4.2-1 16.3/4.3 Instrumentation 16.3/4.3-1
. 16.3/4.4 Reactor Coolant System 16.3/4.4-1 16.3/4.5 Emergency Core Cooling Systems 16.3/4.5-1 16.3/4.6 Containment Systems 16.3/4.6-1 16.3/4.7 Plant Systems 16.3/4.7-1 16.3/4.8 Electrical Power Systems 16.3/4.8-1 16.3/4.9 Refueling Operations 16.3/4.9-1 16.3/4.10 Special Test Exceptions 16.3/4.10-1 7
O 16.3/4-i through 16.3/4-xvii /16.3/4-xviii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 l
16.3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE f] REQUIREMENTS 36.3/4.0 Applicability Applicant to provide t
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 16.3.4.1 Reactivity Control Systems Applicant to provide .
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O 16.3/4 4 neector coo 1e e svetem Applicant to provide O
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O- SECTION 16.B3/4 CONTENTS Section Title Page 16.3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS- 16.B3/4.0-1 16.B3/4.0 , Applicability 16.B3/4.0-1 16.B3/4.1 Reactivity Control Systems 16.B3/4.1-1 16.B3/4.2 Power Distribution Limits 16.B3/4.2-1 16.B3/4.3 Instrumentation 16.B3/4.3-1 16.B3/4.4 Reactor Coolant System 16.B3/4.4-1 16.B3/4.5 Emergency Core Cooling 16.B3/4.5-1 16.B3/4.6 Containment Systems 16.B3/4.6-1 16.B3/4.7 Plant Systems 16.B3/4.7-1
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- - GESSAR II 22A7007 i .238 NUCLEAR ISLAND Rev. 5 SECTION 16.5 CONTENTS Section Title Page 4
16.5 DESIGN FEATURES 16.5.1-1 .
1' 16.5.1 Site 16.5.1-1 16.5.2 , Containment 16.5.2-1
! 16.5.3 Reactor Core -16.5.3-1
! 16.5.4 Reactor Coolant System 16.5.4-l' !
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, 16.6.5 Review and Audit 16.6.1-1 16.6.6 Reportable Occurrence Action 16.6.1-1 16.6.7 Safety Limit Violation 16.6.1-1 16.6.8 Procedures 16.6.1-1 16.6.9 Reporting Requirements 16.6.1-1 16.6.10 Record Retention 16.6.1-1 i
() 16.6.11 16.6.12 Radiation Protection Program High Radiation Area 16.6.1-1 16.6.1-1 4
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GESSAR II 22A7007
(- 238 NUCLEAR ISLAND Rev. 5 4
19.1.2 Chapter 2 - Question / Response Index
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- 1. Subsections shown in parentheses reference the co~rresponding Chapter 19 subsection which details the answer to the question.
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, 2. Darrell G. Eisenhut to Glenn G. Sherwood, " Acceptance Review 4
- of Application for Final Design Approval for 238 Nuclear j Island," December 9, 1981.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O '19.1.3 Chapter 3 - Question / Response Index NRC GESSAR II GESSAR II NRC Question Question y Revision Transmittal Number Number Disposition Number Note 2 210.1 3.1 Subsections 4 n 3.6.1.1.4, 3.6.2.2.1, n 3.6.2.3.1 and' 3.6.2.3.2.2 210.2 3.2 Tables 3.9-11 and 3.9-12 220.1 3.3 Section 3.7.2.6 220.2 3.4 Subsection 19.3.3.3 220.3 3.5 Subsection 19.3.3.5 220.4 3.6 Table 3.8-3 and Sub-section 19.3.3.6 220.5 3.7 Subsection 3.8.2.5 220.6 3.8 Subsection 19.3.3.8 220.7 3.9 Subsection 19.3.3.9 220.8 3.10 Subsection 19.3.3.10 s, 441.2 3.11 Subsection 19.3.3.11 4- -
441.3 3.12 Response in 9/82*
441.4 3.13 Response in 9/82*
441.5 3.14 Response in 9/82*
-/ '
441.6 3.15 Response in 9/82*
441.7 3.16 Response in 9/82*
441.8 3.17 Response in 9/82*
441.9 3.18 Response in 9/82*
441.10 3.19 Subsection 19.3.3.19 4 441.11 3.20 Response in 9/82*
441.12 3.21 Response in 9/82*
441.13 3.22 Response in 9/82*
441.14 3.23 Response in 9/82*
441.15 3.24 Subsection 3A.l.2 4 441.16 3.25 Subsection 3A.l.2 4 441.17 3.26 Response in 9/82*
441.18 3.27 Response in 9/82*
441.19 3.28 Response in 9/82*
441.20 3.29 Response in 9/82*
441.21 3.30 Response in 9/82*
241.22 3.31 Subsection 19.3.3.31 5 241.23 3.32 Response in 9/82*
241.24 3.33 Response in 9/82* '
241.25 3.34 Subsection 19.3.3.34 5 v 241.26 3.35 Subsection 3A.5.2 5 Note 2 251.1 3.36 Subsection 3.5.1.3 4
() *Geotechnical 19.1.3-1 l
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 19.1.3 Chapter 3 - Question / Response Index (Continued)
NRC GESSAR II GESSAR II NRC Question Question y Revision Transmittal Number Number Disposition Number Note 2 270.1 3.37 Tables 3.11-2 5 through 3.11-9 270.2 3.38 Subsection 3.11.4 5 270.3 3.39 Subsection 3.11.2.1.3 5 270.4 3.40 Subsection 3.11.2.1.1 5 Note 2 371.1 3.41 Table 3.10-1 4
~
- Environmental' Qualification _
Chapter - Question / Response Index Notes
- 1. Subsections shown in parentheses reference the corresponding Chapter 19 subsection which details the answer to the question, ggg
- 2. Darrell G. Eisenhut to Glenn G. Sherwood, " Acceptance Review of Application for Final Design Approval for 238 Nuclear Island," December 9, 1981.
- 3. See Section 3BO.1 for Appendix 3B Question / Response Index. ,
O 19.1.3-2
GESSAR II 22A7001 238 NUCLEAR ISLAND Pev. 5 19.1.6 Chapter 6 - Question / Response Index [
NRC GESSAR II GESSAR II ,
NRC Question Question y Revision Transmittal Number s Number Disposition Number
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- 1. Subsectionsshowninparenthesesreferencethecorrespondi$g Chapter 19 subsection which details the answd'r to the question. -
- 2. Darrell G. Eisenhut to Glenn G. Sherwood, Acceptance Review of Application for Final Design Approval for 238 Nuclear Island," December 9, 1981. - -
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
() 19.1.7 Chapter 7 - Question / Response Index NRC GESSAR II GESSAR II NRC Question Question y Revision Transmittal Number Number Disposition Number i
Note 2 420.1 7.1 Subsection 19.3.7.1 4 Note 2 420.2 7.2 Subsection 19.3.7.2 4 ,
Note 2 420.3 7.3 Response in 9/82 -
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4 Subsection 19.3.7.11 Note 2 420.12 7.12 Subsection 19.3.7.12 4 Chapter 7 - Question / Response Index Notes
- 1. Subsections shown in parentheses reference the corresponding Chapter 19 subsection which details the answer to the question.
- 2. Darrell G. Eisenhut to Glenn G. Sherwood, " Acceptance Review of Application for Final Design Approval for 238 Nuclear Island," December 9, 1981.
9 O
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
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A (FMEA portion in 9/82) 410.1 9.2 Subsection 9.1.2.3.2 4 410.2 9.3 Subsections 9.1.4.2.2 and 19.3.9.3 4 410.3 9.4 Response in 9/82 (FMEA) 410.4 9.5 Subsection 9.1.4.3 4 410.5 9.6 Subsection 9.1.4.5.1 4 410.6 9.7 Subsection 9.2.1.3.1 4 410.7 9.8 Subsection 9.2.6 4 410.8 9.9 Response in 9/82 (FMEA) 410.9 9.10 Subsection 9.4.2.1 4 410.10 9.11 Subsection 9.4.3.1 4 k'_)x 430.2 9.12 Subsection 19.3.9.12 4 -
Note 2 430.3 9.13 Subsection 9.5.8.1 4 Chapter 9 - Question / Response Index Notes
- 1. Subsections shown in parentheses reference the corresponding Chapter 19 subsection which details the answer to the question.
- 2. Darrell G. Eisenhut to Glenn G. Sherwood, " Acceptance Review of Application for Final Design Approval for 230 Nuclear Island," December 9, 1981.
s 19.1.9-1/19.1.9-2
GESSAR II 22A7007
. 238 NUCLEAR ISLAND Rev. 5 l () 19.1.12 Chapter 12 - Question / Response Index
! NRC GESSAR II GESSAR II NRC Question' Question y Revision Transmittal Number Number Disposition Number Note 2 471.1 12.1 Subsection'12.2.1.1 4 Note 2 471.2 12.2 Response in.9/82*
Note 2 471.3 12.3 Subsection 19.3.12.3 4 (Subsections 12.3.2.3 t
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Note 2 471.4 12.4 Subsection 12.3.3 4
- Radiation sources and inventories i
j Chapter 12 - Question / Response Index Notes
() 1. Subsections shown in parentheses reference the corresponding Chapter 19 subsection which details the answer to the i question.
- 2. Darrell G. Eisenhut to Glenn G..Sherwood, " Acceptance Review i of Application for Final Design Approval for 238 Nuclear Island," December 9, 1981.
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i GESSAR II. 22A7007 238 NUCLEAR ISLAND Rev. 5 19.3.3.3 Question / Response 3.3(220.1) i i
QUESTION 3.3 J
Indicate the extent to which the recommendations of Regulatory Guide 1.92 are followed. . (3. 7. 2 6) 4
! RESPONSE 3.3 1
4 The extent to which the recommendations of Regulatory Guide 1.92 _
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
() 19.3.3.31 QUESTION / RESPONSE 3.31 (241.22)
QUESTICN 3.31 Provide velocity and displacement time histories corresponding to input acceleration time histories to illustrate that the time histories used in the analysis are base line corrected. (3A.3.1)
RESPONSE 3.31 Figures 19.3.3.31-1 through 19.3.3.31-6 provide the velocity and displacement time histories corresponding to the input acceleration time histories Hy, H2, and V. _
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() 19.3.3.34 QUESTION / RESPONSE 3.34 (241.25)
QUESTION 3.34 You have made a vertical analysis for two profiles. In Figure 3A-24, you have shown nine response spectra. What do these spectra plots represent? Provide a description and table that relates the response spectra plots to the profiles. (3A.5.2)
RESPONSE 3.34 Figure 3A-24 contains response spectra from both the vertical and some of the horizontal SSI analyses cases. The vertical spectra from the horizontal SSI cases have been deleted from this figure.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5
() 19.3.3.35 QUESTION / RESPONSE 3.35 (241.26)
QUESTION 3.35 You have presented the results of free field and interaction response spectra at basemat level for the OBE only. Provide these results for the SSE also. (3A.5.2)
RESPONSE 3.35 Response to this question is provided in Subsection 3A.5.2.
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- O 19.3.3.37 oUEST1oNeRESeoNSE 3.37 <270.1) !
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] Supply the missing information in Tables 3.11-2, 3.11-3, 3.11-8 I and 3.11-9. (3.11.1 and 3.11.2)
RESPONSE 3.37 4
Where appropriate, "To be provided by Applicant" or " Applicant -
to supply" was added to these tables and Tables 3.11-4 through I 3.11-7. .
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O 19.3.3.38 QUESTION / RESPONSE 3.38 (270.2) 4 QUESTION 3.38 Indicate how the requirements of GDC-50 of Appendix A to 10 CFR Part 50 have been met. (3.11.2)
RESPONSE 3.38 The response to this question is provided in Subsection 3.11.2.1.1 -
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 C (
19.3.3.39 QUESTION / RESPONSE 3.39 (270.3)
QUESTION 3.39 Indicate the extent of compliance with NUREG-0588. (3.11.2)
RESPONSE 3.39 The response to this question is provided in Subsection 3.11.2.1.3. _
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I GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 5 O 19.3.3.40 QUESTION / RESPONSE 3.40 (270.4)
QUESTION 3.40 Supply the missing information in this section. (3.11.4)
RESPONSE 3.40 Where appropriate, " Applicant to supply" was added to Section 3.11.4. _
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