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CHAPTER 1
INTRODUCTION AND GENERAL DESCRIPTION OF PLANT
TABLE OF CONTENTS
Section Title Page
1.0 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 1.1-1
1.1 INTRODUCTION 1.1-1
1.2 GENERAL PLANT DESCRIPTION 1.2-1
1.2.1 PRINCIPAL SITE CHARACTERISTICS 1.2-1 (DRN 01-758, R11-A) 1.2.2 CONCISE PLANT DESCRIPTION 1.2-2 (DRN 01-758, R11-A) 1.3 COMPARISONS 1.3-1
1.3.1 COMPARISONS WITH SIMILAR FACILITY DESIGNS 1.3-1
1.3.2 COMPARISON OF FINAL AND PRELIMINARY INFORMATION 1.3-1
1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4-1
1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.5-1
1.5.1 FRETTING AND VIBRATIONS TESTS OF FUEL ASSEMBLIES 1.5-1
1.5.2 DEPARTURE FROM NUCLEATE BOILING (DNB) TESTING 1.5-1
1.5.3 FUEL ASSEMBLY STRUCTURAL TESTS 1.5-1
1.5.4 FUEL ASSEMBLY FLOW MIXING TESTS 1.5-2
1.5.5 REACTOR FLOW MODEL TESTING AND EVALUATION 1.5-2
1.5.6 FUEL ASSEMBLY FLOW TESTS 1.5-3
1.5.7 CONTROL ELEMENT DRIVE MECHANISM (CEDM) TESTS 1.5-3
1.5.8 DNB IMPROVEMENT 1.5-4
1.5 REFERENCES 1.5-4
1.6 MATERIAL INCORPORATED BY REFERENCE 1.6-1
1.7 ELECTRICAL, INSTRUMENTATION, AND CONTROL DRAWINGS 1.7-1
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CHAPTER 1
TABLE OF CONTENTS (Cont'd)
Section Title Page
1.8 COMPARISON OF WATERFORD 3 DESIGN WITH NRC REGULATORY 1.8-1 GUIDES
1.8.1 INTRODUCTION 1.8-1
1.9 THREE MILE ISLAND - 2 (TMI-2) ACTION PLAN REQUIREMENTS 1.9-1 FOR APPLICANTS FOR AN OPERATING LICENSE
1.9.1 SHIFT TECHNICAL ADVISOR (I.A.1.1) 1.9-1
1.9.2 SHIFT SUPERVISOR ADMINISTRATIVE DUTIES (I.A.1.2) 1.9-1
1.9.3 SHIFT MANNING (I.A.1.3) 1.9-2
1.9.4 IMMEDIATE UPGRADING OF REACTOR OPERATOR AND SENIOR 1.9-2 REACTOR OPERATOR TRAINING AND QUALIFICATIONS (I.A.2.1)
1.9.5 ADMINISTRATION OF TRAINING PROGRAMS (I.A.2.3) 1.9-2
1.9.6 REVISE SCOPE AND CRITERIA FOR LICENSING EXAMINATIONS 1.9-3 (I.A.3.1)
1.9.7 INDEPENDENT SAFETY ENGINEERING GROUP (I.B.1.2) 1.9-3
1.9.8 SHORT-TERM ACCIDENT ANALYSIS AND PROCEDURE REVISION 1.9-4 (I.C.1)
1.9.9 SHIFT RELIEF AND TURNOVER PROCEDURES (I.C.2) 1.9-5
1.9.10 SHIFT SUPERVISOR RESPONSIBILITIES (I.C.3) 1.9-5
1.9.11 CONTROL ROOM ACCESS (I.C.4) 1.9-5
1.9.12 PROCEDURES FOR FEEDBACK OF OPERATING EXPERIENCE TO 1.9-6 PLANT STAFF (I.C.5)
1.9.13 GUIDANCE ON PROCEDURES FOR VERIFYING CORRECT PER- 1.9-7 FORMANCE OF OPERATING ACTIVITIES (I.C.6)
1.9.14 NSSS VENDOR REVIEW OF PROCEDURES (I.C.7) 1.9-7
1.9.15 CONTROL ROOM DESIGN REVIEWS (I.D.1) 1.9-8
1.9.16 PLANT SAFETY PARAMETER DISPLAY SYSTEM (I.D.2) 1.9-8
1-ii
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CHAPTER 1
TABLE OF CONTENTS (Cont'd)
Section Title Page
1.9-17 TRAINING DURING LOW-POWER TESTING (I.G.1) 1.9-9
1.9.18 REACTOR COOLANT SYSTEM VENTS (II.B.1) 1.9-9
1.9.19 DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTAL 1.9-10 QUALIFICATION OF EQUIPMENT FOR SPACES/SYSTEMS WHICH MAY BE USED IN POST-ACCIDENT OPERATIONS (II.B.2)
1.9.20 POST-ACCIDENT SAMPLING CAPABILITY (II.B.3) 1.9-11
1.9.21 TRAINING FOR MITIGATING CORE DAMAGE (II.B.4) 1.9-11
1.9.22 PERFORMANCE TESTING OF BOILING-WATER REACTOR AND 1.9-12 PRESSURIZED-WATER REACTOR RELIEF AND SAFETY VALVES (II.D.1)
1.9.23 DIRECT INDICATION OF RELIEF-AND-SAFETY-VALVE POSITION 1.9-12 (II.D.3)
1.9.24 AUXILIARY FEEDWATER SYSTEM EVALUATION (II.E.1.1) 1.9-13
1.9.25 AUXILIARY FEEDWATER SYSTEM AUTOMATIC INITIATION AND 1.9-14 FLOW INDICATION (II.E.1.2)
1.9.26 EMERGENCY POWER SUPPLY FOR PRESSURIZER HEATERS 1.9-15 (II.E.3.1)
1.9.27 DEDICATED HYDROGEN PENETRATIONS (II.E.4.1) 1.9-16
1.9.28 CONTAINMENT ISOLATION DEPENDABILITY (II.E.4.2) 1.9-17
1.9.29 ADDITIONAL ACCIDENT-MONITORING INSTRUMENTATION 1.9-19 (II.F.1)
1.9.30 INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE 1.9-28 COOLING (II.F.2)
1.9.31 EMERGENCY POWER FOR PRESSURIZER EQUIPMENT (II.G.1) 1.9-28
1.9.32 IE BULLETINS ON MEASURES TO MITIGATE SMALL-BREAK LOCAs 1.9-28 AND LOSS OF FEEDWATER ACCIDENTS (II.K.1) (DRN 01-758, R11-A) 1.9.33 ORDERS ON B&W PLANTS (II.K.2) 1.9-29 (DRN 01-758, R11-A) 1.9.34 FINAL RECOMMENDATIONS OF B&O TASK FORCE (II.K.3) 1.9-31
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CHAPTER 1
TABLE OF CONTENTS (Cont'd)
Section Title Page
1.9.35 EMERGENCY PREPAREDNESS - SHORT TERM (III.A.1.1) 1.9-35
1.9.36 UPGRADE EMERGENCY SUPPORT FACILITIES (III.A.1.2) 1.9-35
1.9.36a IMPROVING LICENSEE EMERGENCY PREPAREDNESS - LONG 1.9-36 TERM (III.A.2)
1.9.37 INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT LIKELY TO 1.9-36 CONTAIN RADIOACTIVE MATERIAL FOR PRESSURIZED-WATER REACTORS AND BOILING-WATER REACTORS (III.D.1.1)
1.9.38 IMPROVED INPLANT IODINE INSTRUMENTATION UNDER 1.9-40 ACCIDENT CONDITIONS (III.D.3.3)
1.9.39 CONTROL-ROOM HABITABILITY REQUIREMENTS (III.D.3.4) 1.9-43 (DRN 01-758, R11-A) 1.9A RESPONSE TO SECTION II.F.2 OF NUREG-0737 1.9A-1 INADEQUATE CORE COOLING INSTRUMENTATION (DRN 01-758, R11-A) 1.9A.1 INTRODUCTION 1.9A-1
1.9A.2 FUNCTIONAL DESCRIPTION OF ICCI 1.9A-2
1.9A.3 ICCI SENSOR DESIGN DESCRIPTION 1.9A-3
1.9A.4 SIGNAL PROCESSING AND DISPLAY 1.9A-7
1.9A.5 SYSTEM VERIFICATION TESTING 1.9A-12
1.9A.6 ICCI SYSTEM QUALIFICATION 1.9A-13
1.9A.7 OPERATING INSTRUCTIONS 1.9A-13
1.9A.8 COMPARISON OF DOCUMENTATION REQUIREMENTS WITH 1.9A-14 THIS REPORT
1.9A.9 SCHEDULE FOR ICCI IMPLEMENTATION 1.9A-14
1.9A REFERENCES 1.9A-15
1A FSAR ACRONYMS 1A-1
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CHAPTER 1
INTRODUCTION AND GENERAL DESCRIPTION OF PLANT
LIST OF TABLES
Table Title (DRN 01-758, R11-A) 1.3-1 PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1 (DRN 01-758, R11-A) 1.7-1 ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS
1.9-1 TMI-RELATED REQUIREMENTS FOR NEW OPERATING LICENSES
1.9-2 TMI INFORMATION REQUIRED FOR CONTROL-ROOM HABITABILITY EVALUATION (TASK ACTION PLAN ITEM III.D.3.4)
1.9-3 CONTAINMENT ISOLATION VALVES PROVIDED WITH CAPABILITY FOR MANUAL OVERRIDE
1.9-4 ADDITIONAL ACCIDENT MONITORING INSTRUMENTATION
1.9A-1 DEFINITION OF ICC EVENT PROGRESSION INTERVALS
1.9A-2 COMPARISON OF ICCI TO DOCUMENTATION REQUIREMENTS OF ITEM II.F.2 OF NUREG-0737
1.9A-3 COMPARISON OF ICCI TO ATTACHMENT 1 OF II.F.2
1.9A-4 COMPARISON OF ICCI TO APPENDIX B OF NUREG-0737
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CHAPTER 1
INTRODUCTION AND GENERAL DESCRIPTION OF PLANT
LIST OF FIGURES
Figure Title
1.2-1 PLOT PLAN
1.2-2 SITE PLAN
1.2-3 GENERAL ARRANGEMENT TURBINE BUILDING GROUND FLOOR - PLAN
1.2-4 GENERAL ARRANGEMENT TURBINE BUILDING MEZZANINE FLOOR - PLAN
1.2-5 GENERAL ARRANGEMENT TURBINE BUILDING OPERATING FLOOR - PLAN
1.2-6 GENERAL ARRANGEMENT TURBINE BUILDING - SECTION
1.2-7 GENERAL ARRANGEMENT TURBINE BUILDING - SECTIONS
1.2-8 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING PLAN EL + 46.00’
1.2-9 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING PLAN EL + 21.00’
1.2-10 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING PLAN EL - 4.00’
1.2-11 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING PLAN EL - 35.00'
1.2-12 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING - SECTION
1.2-13 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING - SECTION (DRN 01-758, R11-A) 1.2-14 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING - (SH. 3) (DRN 01-758, R11-A) 1.2-15 GENERAL ARRANGEMENT FUEL HANDLING BUILDING - PLANS
1.2-16 GENERAL ARRANGEMENT FUEL HANDLING BUILDING - SECTIONS
1.2-17 GENERAL ARRANGEMENT REACTOR BUILDING PLAN - EL + 46.00' (DRN 01-758, R11-A) 1.2-18 GENERAL ARRANGEMENT REACTOR BUILDING - PLAN - EL + 21.00' (DRN 01-758, R11-A) 1.2-19 GENERAL ARRANGEMENT REACTOR BUILDING - PLAN EL - 4.00' (DRN 01-758, R11-A) 1.2-20 GENERAL ARRANGEMENT REACTOR BUILDING - SECTION (SHEET 1)
1.2-21 GENERAL ARRANGEMENT REACTOR BUILDING - SECTION (SH 2) (DRN 01-758, R11-A) 1.2-22 GENERAL ARRANGEMENT REACTOR BUILDING - SECTIONS (DRN 01-758, R11-A) 1.2-23 GENERAL ARRANGEMENT REACTOR AUXILIARY BUILDING - PLANS AND SECTIONS - (SHEET 4) (DRN 01-758, R11-A)
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CHAPTER 1
LIST OF FIGURES (Cont'd)
Figure Title
1.2-24 GENERAL ARRANGEMENT COOLING TOWERS - PLAN
1.2-25 GENERAL ARRANGEMENT COOLING TOWERS - SECTIONS
1.9-1 DELETED
1.9-2 DELETED
1.9-3 BLOCK DIAGRAM, WIDE-RANGE GAS MONITOR
1.9-4 DETECTOR RANGES, WIDE-RANGE GAS MONITOR
1.9A-1 DEFINITION OF INTERVALS IN EVENT PROGRESSION
1.9A-2 ICC INSTRUMENTATION SYSTEM
1.9A-3 HJTC SENSOR
1.9A-4 HJTC SPLIT PROBE DESIGN CONFIGURATION
1.9A-5 HJTC SENSOR AXIAL LOCATIONS
1.9A-6 INADEQUATE CORE COOLING REACTOR VESSEL LEVEL CABLE ROUTING
1.9A-7 CORE EXIT TEMPERATURE MEASUREMENT SCHEME
1.9A-8 CORE EXIT THERMOCOUPLE CORE LOCATIONS
1.9A-9 INADEQUATE CORE COOLING CORE EXIT THERMOCOUPLES CABLE ROUTING
1.9A-10 HJTC LEVEL LOGIC
1.9A-11 HEATER POWER CONTROL LOGIC
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Chapter 1
Section Cross References
Revision 12-B None
Revision 12-C Page 1A-7 ER-W3-2003-0249-000/DRN 03-657 Page 1.9-4
Revision 13 Page 1.9A-6 ER-W3-2003-0681/DRN 03-1872
Revision 13-A
Section 1.1 ER-W3-2000-1018-009/DRN 04-1302
Revision 13-B
Figure 1.2-3 ER-W3-2004-0500/DRN 04-1514
Section 1.7 ER-W3-2004-0502/DRN 04-1444 Table 1.7-1
Revision 14
Section 1.1 ER-W3-2001-1149-000/DRN 03-2054 Section 1.2.1.3 Section 1.2.2.1.1 Section 1.2.2.1.2 Section 1.2.2.9 Section 1.6 Section 1.8.1.46 Section 1.8.1.145 Section 1.9.28 Section 1.9A.1.1 Section 1.9A.2.3 Section 1.9A.3.1 Section 1.9A.3.3 Section 1.9A.4.1.1 Section 1.9A.4.1.2 Section 1.9A.4.1.3 Section 1.9A.6
Figure 1.2-4 ER-W3-2002-0602-001/DRN 03-1184
Figure 1.9A-5 ER-W3-2004-0137-000/DRN 04-1958
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UPDATE REFERENCE LIST
Chapter 1
Section Cross References
Revision 14 Cont’d
Section 1.2.2.2 ER-W3-2004-0276-000/DRN 04-1619 Section 1.8.1.4 Section 1.8.1.24 Section 1.8.1.25 Section 1.8.1.77 Section 1.8.1.183 Section 1.9.28 Section 1.9.37 Section 1.9.39
Revision 14-A
Section 1.9.37 ER-W3-2005-0378-000/DRN 05-1265
Revision 14-B
Figure 1.2-7 ER-W3-2005-0442-000/DRN 06-278
Revision 15
Appendix 1A ER-W3-2006-0210-000/DRN 06-623
Section 1.8.1.46 ER-W3-2006-0258-000/DRN 06-802
Section 1.2.2.1.1 ER-W3-2005-0447-004/DRN 06-1058
Revision 301
Figure 1.2-18 EC-1396
Section 1.8.1.23 EC-1837
Section 1.8.1.45 EC-5000082437
Table 1.9-2 Sh. 1 EC-5000082445
Revision 302
Figure 1.9A-5 EC-6607
Revision 303
Figure 1.2-17 EC-8039
Revision 304
Figure 1.9A-5 EC-10453
Section 1.5.2 EC-13881 Section 1.6
Section 1.2.2.3.1 EC-15702
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UPDATE REFERENCE LIST
Chapter 1
Section Cross References
Revision 304 Cont’d
Section 1.2.2.6 EC-16212
Section 1.9A.3.3 EC-18688 Section 1.9A.4 Section 1.9A.4.1.3 Figure 1.9A-8
Revision 305
Figure 1.2-4 EC-8038
Section 1.6 EC-19087 Section 1.8.45
Section 1.8.1.133 EC-26965
Revision 306
Section 1.9.29 EC-12329 Section 1.9A.4 Section 1.9A.4.2 Table 1.9A-3 Table 1.9A-4 (Sheet 1)
Appendix 1A EC-14275 Section 1.2.2.6
Figure 1.2-18 EC-17580
Revision 307
Section 1.2.2.1.2 EC-1020
Table 1.7-1 EC-2800 Section 1.8.1.31
Section 1.8.1.83 EC-8458
Figure 1.2-17 EC-27161 Figure 1.2-20
Section 1.2.2.1.1 EC-30663
Section 1.9.24 EC-33720
Section 1.8.1.40 EC-40281
Revision 308
Section 1.8.1.9 LBDRC 14-010
Section 1.8.1.143 EC 47424
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Chapter 1
Section Cross References
Revision 308 Cont’d
Section 1.9.2 LBDCR 13-015 Section 1.9.7 Section 1.9.10
Revision 308A
Section 1.9 LBDCR 15-028
Revision 309
Section 1.9.13 LBDCR 15-027 Section 1.9.17 LBDCR 13-014
Revision 310
Section 1.2.1.1 LBDCR 17-003 Section 1.2.2.4.6 EC-29861 Figure 1.2-9 LBDCR 14-029 Figure 1.2-11 Figure 1.2-18 LBDCR 14-029/16-036 Section 1.3.2.9.2 LBDCR 16-012 Section 1.8.1.39 Section 1.8.1.75 Section 1.8.1.88 LBDCR 16-054 Section 1.8.1.199 LBDCR 17-007
Revision 311
Figure 1.2-18 LBDCR 14-012 Figure 1.2-24 LBDCR 16-043 Figure 1.2-25
Figure 1.2-24 LBDCR 16-044 Figure 1.2-25
Figure 1.2-5 LBDCR 18-031
Section 1.8 LBDCR 19-013
Section 1.8.1.9 LBDCR 18-018
Figure 1.9A-5 LBDCR 18-004
1-xi Revision 311 (9/19)
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1.0 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT
1.1 INTRODUCTION
This Final Safety Analysis Report is submitted in support of an application by the Louisiana Power and Light Company for a license to operate a nuclear powered Electric generating unit designated Waterford Steam Electric Station Unit No. 3. The unit is located on the west (right descending) bank of the Mississippi River in St. Charles Parish, near the town of Taft, Louisiana.
The Nuclear Steam Supply System (NSSS) is a pressurized water reactor designed by Combustion Engineering Incorporated. The containment structure is comprised of a steel containment vessel surrounded by a reinforced concrete Shield Building and was designed by Ebasco Services Incorporated.
(DRN 02-691, R12) The Waterford 3 Facility Operating License was issued on March 16, 1985 for a reactor core power level not in excess of 3390 megawatts thermal (MWt). The rated NSSS thermal power level of 3410 MWt, included a 20 MWt contribution from the reactor coolant pumps. The design thermal power level is 3560 MWt, the maximum expected output of the core. This is the basis for the design of the balance of plant and related facilities, including the major systems and components, the engineered safety features, and for site evaluation calculation (see Chapter 15 for details). The corresponding net design electrical outputs are 1104 MWe and 1151 MWe for the 3410 MWt rated power level and 3560 MWt design power level respectively. The corresponding gross electrical outputs are 1153 MWe and 1200 MWe, respectively.
The Facility Operating License was amended on March 29, 2002 to increase the reactor core power level from 3390 MWt to 3441 MWt. The increase in design net electrical output due to the core power level increase is approximately 16 MWe.
(DRN 03-2054, R14) The Facility Operating License was amended starting with Operating Cycle 14 to increase the reactor core power level from 3441 MWt to 3716 MWt. The expected increase in design net electrical output due to the core power level increase is approximately 68 MWe. (DRN 02-691, R12) (DRN 04-1302, R13-A) (DRN 04-1302. R13-A; 03-2054, R14)
Regulatory Guide 1.70, Revision 2, September 1975 was used as a format guide for the Waterford 3 FSAR.
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1.2 GENERAL PLANT DESCRIPTION
1.2.1 PRINCIPAL SITE CHARACTERISTICS
1.2.1.1 Location and Population
-003, R310) The Waterford 3 site is located on the west (right descending) bank of the Mississippi River near Taft, Louisiana in the northwest portion of St. Charles Parish. About three miles westward is the eastern boundary of St. John the Baptist Parish. The coordinates for the reactor are 29o 59' 42" north latitude and 90o 28'16" west longitude. The UTM coordinates are 3320744 meters north, and 743963 meters east. -003, R310) Kenner, the nearest population center is 13 miles east of the site. Approximately 25 miles east-southeast of the site is the city of New Orleans, and approximately 50 miles north-northwest is the city of Baton Rouge. The exclusion radius is taken as 915 meters, and the low population zone is a two mile radius.
1.2.1.2 Geography and Hydrology
The site is located on the right descending bank of the Mississippi River near Taft, Louisiana. It consists of over 3,000 acres of flat land extending from the Mississippi River to the St. Charles Drainage Canal. The site includes about 7500 feet of river frontage. About 3,000 feet back from State Road 18, which runs adjacent to the levee, the Missouri Pacific Railway crosses the width of the property. The plant area is raised to a final grade of +17.5 ft.MSL around the Nuclear Plant Island Structure, and +14.5 ft. MSL around the Turbine Building.
Flood protection in the vicinity of the site includes levees, bypass channels, and channel stabilization that can effectively confine flood flows except for very severe floods. Structures housing safety-related equipment are flood protected to elevation +30 ft.MSL.
1.2.1.3 Meteorology (DRN 03-2054, R14) The climate of southeastern Louisiana is classified as humid subtropical and is characterized by mild, humid winters and hot, humid summers. Daily maximum summer temperatures are generally around 95F while winter temperatures normally lie between 41 and 69F. The region's rainy season extends from mid-December to mid-March. Measurable precipitation occurs on about one-third of the days during this period. Snowfall amounts are very light with the snow usually melting as it falls. (DRN 03-2054, R14)
Thunderstorms with damaging winds and hail are relatively infrequent. A few of the more severe thunderstorms however, will generate tornadoes. The probability of a tornado striking the site is discussed in Section 2.3.
During the period 1871-1977, 55 tropical storms and hurricanes passed within 100 nautical miles of the site.
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Wind direction varies widely at the site with an average speed of about nine mph. Calms occur approximately 12 percent of the time. The average wind speed during inversion conditions is about three mph.
Stability at the site is Pasquill type E,F,or G, 56 percent of the time.
1.2.1.4 Geology and Seismology
The site has a uniform stratigraphy i.e., no salt domes or possibility of local faulting, no possible surface expression of known or hypothetical faults. The structures are conservatively designed and built with respect to geological considerations. The site ground accelerations for the Safe Shutdown Earthquake (SSE) and Operating Basis Earthquake (OBE) are 0.10g and 0.05g, respectively.
1.2.1.5 Nearby Industry and Commerce
The area adjacent to the site is moderately industrialized. Both banks of the Mississippi River near the site are lined with industrial facilities, primarily chemical plants. The Agrico Chemicals Co. is adjacent to and downstream of the Waterford 3 site. Next to Agrico Chemicals Co. is the Occidental Chemical Corporation and farther downstream is the Union Carbide Company. Two Shell Oil Company plants, a chemical plant and a refinery, are located on the opposite bank from Waterford 3, approximately 1.5 miles downstream. The Mississippi River is used extensively for commercial traffic and municipal, and industrial water use.
Site and plot plans are provided in Figures 1.2-1 and 1.2-2.
For further information see Chapter 2.
1.2.2 CONCISE PLANT DESCRIPTION
1.2.2.1 Reactor and Reactor Coolant System
1.2.2.1.1 Reactor (DRN 03-2054, R14) The pressurized water reactor was designed for an initial core thermal power output of 3390 megawatts. The design core thermal power output is increased to 3716 megawatts starting at Cycle 14. (DRN 03-2054, R14)
(DRN 06-1058, R15; EC-30663, R307) The reactor core is fueled with uranium dioxide pellets enclosed in zircaloy or ZirloTM1 tubes pressurized with helium and fitted with welded end plugs. The tubes are fabricated into assemblies in which end fittings prevent axial motion and spacer grids prevent lateral motion of the tubes. The control element assemblies (CEAS) consist of Ni-Cr-Fe alloy clad boron carbide absorber rods, guided by tubes in the fuel assembly. The core consists of 217 fuel assemblies with multiple U-235 enrichment. (DRN 06-1058, R15; EC-30663, R307)
Fuel rod clad is designed to maintain cladding integrity throughout fuel life. Fission gas release within the rods and other factors affecting design life are considered for the maximum expected exposure.
(EC-30663, R307) 1 Zirlo, Optimized Zirlo, and Low Tin Zirlo are trademarks or registered trademarks of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners. (EC-30663, R307)
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The reactor and control systems are designed so that any xenon transients will be adequately damped.
The CEAs are capable of holding the core subcritical at hot zero power conditions with margin following a trip even with the most reactive CEA stuck in the fully withdrawn position.
The combined response of the fuel temperature coefficient, the moderator temperature coefficient, the moderator void coefficient and the moderator pressure coefficient to an increase in reactor thermal power is a decrease in reactivity. In addition, the reactor power transient remains bounded and damped in response to any expected changes in any operating variable.
The reactor, in conjunction with its protective systems is designed to safely accommodate the anticipated operational occurrences.
See Chapter 4 for further information.
1.2.2.1.2 Reactor Coolant System
The Reactor Coolant System (RCS) is arranged as two closed loops connected in parallel to the reactor vessel. Each loop consists of one 42 in. ID outlet (hot) pipe, one steam generator, two 30 in. ID inlet (cold) pipes and two pumps. An electrically heated pressurizer is connected to one of the loops and a safety injection line is connected to each of the four inlet legs. The RCS operates at a nominal pressure of 2,250 psia.
(EC-1020, R307) The reactor vessel is fabricated from SA-533, Grade B steel, clad with stainless steel. The replacement reactor vessel closure head is fabricated from SA-508, Grade 3, Class 1 material. The design of the vessel and its internals is such that for reactor operation at design power and an 80 percent capacity factor, the vessel fluence greater than one Mev at the vessel wall will not exceed 3.68 x 1019 n/cm2 over the 40-year design life of the vessel. (EC-1020, R307)
(DRN 03-2054, R14) The two steam generators are vertical shell and U-tube units. The steam generated in the shell side of the steam generator flows upward through moisture separators which reduce its moisture content to less than 0.25 percent. All RCS internal surfaces are either stainless steel or Ni-Cr-Fe alloy in order to maintain reactor coolant purity. (DRN 03-2054, R14)
The reactor coolant is circulated by four electric-motor-driven, single-suction centrifugal pumps. The pump shafts are sealed by mechanical seals. The seal performance is monitored by pressure and temperature sensing devices in the seal system.
The RCS is designed and constructed to maintain its integrity throughout the plant life. Appropriate means of test and inspection are provided.
See Chapter 5 for further information.
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1.2.2.2 Engineered Safety Features
(DRN 04-1619, R14) The plant design incorporates redundant engineered safety features (ESF). These systems, along with the containment, insure that the offsite radiological consequences following any postulated loss-of-coolant accident (LOCA) up to and including a double-ended break of the largest reactor coolant pipe will not exceed the guidelines of 10CFR50.67. The systems also insure that the guidelines of 10CFR50, Appendix K, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Power Reactors", are satisfied, based upon analytical methods, assumptions and procedures accepted by the NRC. The ESF include: (a) independent redundant systems (Containment Cooling System and Containment Spray System) to remove heat from and reduce the pressure in the containment in order to maintain containment integrity, (b) a Safety Injection System to limit fuel and cladding damage to an amount which will not interfere with adequate emergency core cooling and to limit metal-water reactions to negligible amounts, (c) a Containment Isolation System to minimize post-LOCA radiological effects offsite, (d) a Combustible Gas Control System to maintain safe post-LOCA hydrogen concentration within the containment (e) Habitability Systems to insure control room habitability following a LOCA, and (f) a Shield Building Ventilation System to limit annulus pressure, and to control and filter releases post-LOCA. (DRN 04-1619, R14)
See Chapter 6 for further information.
1.2.2.3 Instrumentation and Controls
1.2.2.3.1 Controls (EC-15702, R304) The Reactor Control System is used for startup and shutdown of the reactor and for adjustment of the reactor power in response to turbine load demand. The Nuclear Steam Supply System (NSSS) is capable of following a ramp change from 15 percent to 100 percent power at a rate of up to five percent per minute and at greater rates over smaller load change increments up to a step change of 10 percent. During a maneuver, compensation must be provided for the changes in reactivity associated with both changes in power level (power defect) and changes in transient xenon level which result from the change in power level. The average temperature control program provides a reference temperature which is a function of power. This temperature is compared with the existing average reactor coolant temperature. If the temperature is different, core reactivity is adjusted until the difference is within the prescribed control band. Regulation of the average reactor coolant temperature in accordance with this program maintains the secondary steam pressure within operating limits and matches reactor power to load demand. The mechanisms by which the necessary reactivity compensation is provided are CEA position, boron concentration and primary coolant temperature. (EC-15702, R304)
The reactor is controlled by a combination of CEAs and dissolved boric acid in the reactor coolant. Boric acid is used for reactivity changes associated with large but gradual changes in average coolant temperature, xenon concentration and fuel burnup.
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Additions of boric acid also provide an increased shutdown margin during the initial fuel loading and subsequent refuelings. The boric acid solution is prepared and stored at a temperature sufficiently high to prevent precipitation.
CEA movement provides changes in reactivity for shutdown or power changes. The CEAs are actuated by control element drive mechanisms mounted on the reactor vessel head. The control element drive mechanisms are designed to permit rapid insertion of the CEAs into the reactor core by gravity. CEA motion can be initiated manually or automatically.
The Core Operating Limit Supervisory System (COLSS) functions to monitor selected parameters, to provide an on-line calculation of margin to a Limiting Condition for Operation (LCO), and to actuate an alarm when a LCO is reached.
The pressure in the RCS is controlled by regulating the temperature of the coolant in the pressurizer, where steam and water are held in thermal equilibrium. Steam is formed by the pressurizer heaters or condensed by the pressurizer spray to reduce variations caused by expansion and contraction of the reactor coolant due to system temperature changes.
Overpressure protection is provided by safety valves connected to the pressurizer and designed in accordance with ASME Code, Section III. The discharge from the pressurizer safety valves is released under water in the quench tank, where it is condensed and cooled. Overpressure protection for the tank is provided by a rupture disc which relieves to containment.
1.2.2.3.2 Instrumentation
The nuclear instrumentation includes out-of-core and in-core flux detectors. Eight independent channels of out-of-core nuclear instrumentation monitor the fission process. Two channels are used to monitor the reactor from startup through full power; four channels are used to monitor the reactor from routine startup neutron flux levels to 200 percent power and are used to initiate a reactor shutdown in the event of high linear or logarithmic power; and two are used in the automatic control system to regulate the reactor in response to turbine demand. The in-core monitors provide information on neutron flux distribution.
The reactor parameters are maintained within acceptable limits by the inherent negative feedback characteristics of the reactor, by the CEAS, by boric acid dissolved in the moderator, and by the operating procedures. In addition, in order to preclude unsafe conditions for plant equipment or personnel, the RPS is provided. The RPS consists of sensors, calculators, logic and other equipment necessary to monitor selected Nuclear Steam Supply System (NSSS) conditions and to effect reliable and rapid reactor shut- down (reactor trip) if any or a combination of the monitored conditions approach specified safety system settings. The system's functions are to protect the core fuel design limits and reactor coolant pressure boundary for anticipated operational occurrences and also to provide assistance in limiting conditions for certain accidents. Four measurement channels with electrical and physical separation are provided for each parameter used in the direction generation of trip signals, with the exception of control element assembly (CEA) position. A two out of four coincidence of like trip signals is required to generate a reactor trip signal. The use of four channels allows bypassing of one channel for maintenance while maintaining a two out of three channel trip. The reactor trip signal deenergizes the control element drive mechanism (CEDM) coils, allowing all CEAs to drop into the core.
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Four independent core protection calculators (CPCS) are provided, one in each protection channel. Calculation of departure from nucleate boiling ratio (DNBR) and local power density is performed in each CPC, utilizing the input signals described below. The DNBR and local power density so calculated are compared with trip set-points for initiation of a low DNBR trip and the high local power density trip.
Two independent CEA calculators are provided as part of the CPC System to calculate individual CEA deviations from the position of the other CEAs in their subgroup.
Each CPC receives the following inputs: core inlet and outlet temperature, pressurizer pressure, reactor coolant pump speed, excore nuclear instrumentation flux power (each subchannel from the safety channel), selected CEA position, and CEA subgroup deviation from the CEA calculators. Input signals are conditioned and processed.
Additional temperature, pressure, flow and liquid level monitoring is provided, as required, to keep the operating personnel informed of plant conditions, and to provide information from which plant processes can be evaluated and/or regulated.
The plant gaseous and liquid effluents are monitored for radioactivity. Activity levels are displayed and off- normal values are annunciated. Area monitoring stations are provided to measure radioactivity at selected locations in the plant.
See Chapter 7 for further information.
1.2.2.4 Electric Power
Waterford 3 generates power at a nominal 25 kV. This is transformed up to 230 kV and enters the 230 kV switchyard through two overhead tie lines. Two start-up transformers, each supplied from one of the two overhead tie lines provide power for start-up, shutdown, reserve full load operation and preferred emergency shutdown service to the 6.9 kV and 4.16 kV auxiliary system buses. While the unit is in normal operation, these buses are normally supplied by two auxiliary transformers connected to the main generator 25 kV bus.
Redundant sources of offsite power are provided by seven separate transmission lines connected to the 230 kV switchyard. Any one of these lines together with either of the tie lines and its start-up transformer is capable of supplying the total emergency power requirements to ensure that no single failure of any active component can prevent a safe and orderly shutdown.
Redundant sources of onsite power are provided by two diesel generators, either of which is capable of supplying sufficient engineered safety features (ESF) loads to ensure safe shutdown and maintenance in a safe condition in the event of complete loss of offsite power.
The ESF redundant systems have been electrically and physically designed and segregated so that a single electrical fault or a single credible event will not cause loss of power to both sets of redundant essential electrical components.
See Chapter 8 for further information.
1.2-6
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1.2.2.5 Steam and Power Conversion System
The Steam and Power Conversion System removes heat energy from the reactor coolant in two U-tube steam generators, and converts the steam into electrical energy by means of a turbine-generator. The unusable heat in the steam cycle is transferred to the main condenser for rejection by the Circulating Water System. The resulting condensate is then deaerated, heated through feedwater heaters and returned to the steam generators as feedwater.
The main turbine is a Westinghouse Electric Corporation 1800 rpm, tandem-compound, six flow exhaust unit with 40 in. last stage blades. Moisture separators and reheaters dry and superheat the steam between the high and low pressure elements of the turbine.
The main condenser is a single-pass, three shell, single pressure type with divided water boxes. The tubes in each shell are oriented transverse to the turbine shaft. Cooling water for the condenser is provided by the Circulating Water System from the Mississippi River.
Other components of the Steam and Power Conversion System are the main steam supply piping, Steam Bypass System, three motor driven condensate pumps, three strings of five stage low-pressure feedwater heaters, two turbine driven feedwater pumps, three strings of one stage high pressure feedwater heaters, the Steam Generator Blowdown System, and Emergency Feedwater System.
The Emergency Feedwater System (EFS) supplies condensate to the steam generators following the loss of normal feedwater. The EFS also provides water to the unaffected steam generator following a postulated main steam or feedwater line break.
In case of a turbine trip, the Steam Bypass System passes steam directly to the condenser thereby dissipating the stored energy in the reactor coolant and nuclear fuel. There are also the spring loaded main steam safety valves which discharge to the atmosphere, providing overpressure protection for the Main Steam System.
See Chapter 10 for additional information.
1.2.2.6 Fuel Storage and Handling
New fuel assemblies are normally stored in vertical racks of the spent fuel Pool. (EC-16212, R304; EC-14275, R306) Irradiated fuel assemblies are stored in the spent fuel pool or at the Independent Spent Fuel Storage Installation (ISFSI). The stainless steel lined, reinforced concrete spent fuel pool provides storage for up to 1849 assemblies. Adequate spacing in the spent fuel storage pool precludes criticality; the supporting analysis takes no credit for the boron in the pool water. (EC-16212, R304) The ISFSI, shown on Figure 1.2-1, consists of a concrete pad with space for 72 natural convection air- cooled HI-STORM shielded dry casks, each capable of storing 32 spent fuel assemblies in a welded multipurpose container. The ISFSI is located in the plant protected area. (EC-14275, R306)
New fuel can also be stored in a separate dry fuel storage vault. That vault has vertical rack space for 80 assemblies. New fuel assembly spacing and vault construction precludes criticality.
Cooling and purification equipment is provided for the spent fuel pool cooling water.
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(DRN 01-758) The Fuel Handling System provides for the safe handling of fuel assemblies and control element assemblies (CEAS) and for the required assembly, disassembly and storage of reactor internals. This system includes a refueling machine located inside the containment above the refueling pool, the fuel handling crane, fuel handling tools, the fuel transfer carriage, the upending machine, CEA change mechanism, new fuel elevator, fuel inspection stand, the fuel transfer tube, a fuel handling machine in the spent fuel storage room, and various devices used for handling and storing the reactor vessel head and internals. (DRN 01-758) See Section 9.1 for further information.
1.2.2.7 Cooling Water and Other Auxiliary Systems
1.2.2.7.1 Circulating Water System
The Circulating Water System provides a heat sink with sufficient capacity to remove the heat rejected in the main condenser and Turbine Building Closed Cooling Water System during normal operation. River water is pumped from the intake structure to the tube side of the main condensers and turbine building closed cooling water heat exchangers by the circulating water pumps. Water from the condensers and the heat exchangers is discharged through a system to a discharge structure which discharges into the river downstream of the intake structure.
See Subsection 10.4.5 for further information.
1.2.2.7.2 Component Cooling Water Systems
The Component Cooling Water System (CCWS) is the ultimate heat sink for the plant. It is designed to remove heat from the reactor coolant and the auxiliary systems during normal operation, shutdown, or emergency shutdown following a Loss of Coolant Accident (LOCA). In the CCWS, cooling water is pumped through the dry cooling towers and the tube side of the component cooling heat exchangers, through the components being cooled, and back to the pumps.
The Auxiliary Component Cooling System (ACCS) removes heat, if required, from the CCWS via the component cooling heat exchangers during normal operation, shutdown, or post-LOCA. In the ACCS the cooling water is pumped through the shell side of the component cooling heat exchangers, where it removes heat from the CCWS, and rejects it to the atmosphere via the wet cooling towers.
There are two redundant, independent, full capacity CCWS trains. Each CCWS is provided with an ACCS loop.
See Section 9.2 for a discussion of the CCWS.
1.2-8 Revision 11-A (02/02)
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1.2.2.7.3 Chemical and Volume Control System
A Chemical and Volume Control System (CVCS) controls the purity and chemistry of the reactor coolant. Part of the reactor coolant is bypassed through the CVCS, via regenerative and letdown heat exchangers, a filter and ion exchangers before being sprayed into the volume control tank. The charging pumps take suction from this tank and pump the coolant back into the Reactor Coolant System.
See Subsection 9.3.4 for further information.
1.2.2.7.4 Shutdown Cooling System (DRN 01-758, R11-A) The Shutdown Cooling System (SDCS) reduces the temperature of the reactor coolant from 350°F to the refueling temperature, removes decay heat during normal shutdown, and removes heat from the Safety Injection System sump water via the shutdown cooling heat exchangers following a LOCA. (DRN 01-758, R11-A) During shutdown cooling, a portion of the reactor coolant, via the shutdown cooling lines and low pressure safety injection system pumps, is cooled through two shutdown heat exchangers. The control valves and bypass lines are used to control the plant cooldown rate.
See Subsection 9.3.6 for further information.
1.2.2.7.5 Compressed Air System
The Compressed Air System is provided to supply properly conditioned compressed air required to operate pneumatic instruments and controls, periodically pressurize containment penetrations for leak detection, operate containment isolation valves and perform normal plant maintenance. it consists of the Instrument Air System which supplies the various air operated valves, pneumatic instruments and controls and the Station Air System which supplies various outlets throughout the plant.
Redundancy is provided by multiple compressor units and a cross-connection between the Instrument and Station Air Systems. In case of loss of instrument air, all safety related pneumatically operated devices in the plant are designed to fail in a position which would allow safe shutdown. Where safety class valves are required to operate, accumulators are provided.
See Subsection 9.3.1 for further information.
1.2.2.7.6 Demineralized Makeup Water System (EC-29861, R310) In the Makeup Water System, parish water is first filtered, then demineralized and stored in a condensate storage tank and primary storage tank for use as makeup for the plant processes. (EC-29861, R310) See Subsection 9.2.3 for further information.
1.2-9 Revision 310 (12/17)
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1.2.2.8 Radioactive Waste Management Systems
The Boron Management and Waste Management Systems (BMS and WMS) provide the means for controlled handling, storage and disposal of liquid, gaseous and solid wastes. In addition, the BMS provides the mechanism for reconcentrating and recovering dissolved boron from the liquid effluent for reuse in the plant. (DRN 00-1053, R11-A; 00-803, R11-B,) Liquid effluent from the RCS first passes through the purification filter in the CVCS. It is then processed in the BMS by successively passing through the holdup tanks, filters, and ion exchangers. These operations remove the radioactive material and concentrate the boric acid for reuse or drumming. All other radioactive liquid wastes are processed in the WMS for release to the environment or drumming. All liquid wastes are sampled prior to release. The waste release rates are as low as reasonably achievable and within the guidelines and limits for waste release established by 10CFR20 and 10CFR50, Appendix I. (DRN 00-1053, R11-A; 00-803, R11-B) All solid wastes are stored in suitable containers for ultimate offsite disposal in accordance with applicable regulations.
Waste gases are either collected in the gas surge header or filtered and released to the atmosphere via the gas collection header, depending on expected activity level. High activity gases are collected in the gas surge header and compressed into gas decay tanks. The waste gas held in the gas decay tanks is released to the plant vent after sampling. The tank contents are released at rates well within the limits established by 10CFR20 and 10CFR50, Appendix I.
See Chapter 11 for further information.
1.2.2.9 General Arrangement of Major Structures and Equipment
(DRN 03-2054, R14) The general arrangement of major structures and equipment is shown in Figures 1.2-3 through 1.2-25. (DRN 03-2054, R14)
1.2-10 Revision 14 (12/05)
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ç (DRN 01-415)
Figure 1.2-1 has been incorporated by reference in accordance with NEI 98-03.
Figure information can be found in Drawing G127.
Ï (DRN 01-415)
Revision 11 (05/01) WSES-FSAR-UNIT-3
ç (DRN 01-415)
Figure 1.2-2 has been incorporated by reference in accordance with NEI 98-03.
Figure information can be found in Drawing G128.
Ï (DRN 01-415)
Revision 11 (05/01) WSES-FSAR-UNIT-3
G(DRN 04-1514, R13-B)
Figure 1.2-3 has been incorporated by reference in accordance with NEI 98-03.
Figure information can be found in Drawing G-129.
0(DRN 04-1514, R13-B)
Revision 13-B (01/05)
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→ (LBDCR 18-031, R311)
Figure 1.2-5 has been incorporated by reference in accordance with NEI 98-03
Figure information can be found in Drawing G131
← (LBDCR 18-031, R311)
Revision 311 (9/19)
WSES-FSAR-UNIT-3
ç (DRN 01-415)
Figure 1.2-8 has been incorporated by reference in accordance with NEI 98-03.
Figure information can be found in Drawing G-134.
Ï (DRN 01-415)
Revision 11 (05/01) WSES-FSAR-UNIT-3
( DRN 01-415; LBDCR 14-029, R310)
Figure 1.2-9 has been incorporated by reference in accordance with NEI 98-03.
Figure information can be found in Drawing G-135.
( DRN 01-415; LBDCR 14-029, R310)
Revision 310 (12/17)
WSES-FSAR-UNIT-3
ç (DRN 01-415)
Figure 1.2-10 has been incorporated by reference in accordance with NEI 98-03.
Figure information can be found in Drawing G-136.
Ï (DRN 01-415)
Revision 11 (05/01) WSES-FSAR-UNIT-3
(DRN 01-415; LBDCR 14-029, R310)
Figure 1.2-11 has been incorporated by reference in accordance with NEI 98-03.
Figure information can be found in Drawing G137.
(DRN 01-415; LBDCR 14-029, R310)
Revision 310 (12/17)
WSES-FSAR-UNIT-3
ç (DRN 01-415)
Figure 1.2-12 has been incorporated by reference in accordance with NEI 98-03.
Figure information can be found in Drawing G138.
Ï (DRN 01-415)
Revision 11 (05/01) WSES-FSAR-UNIT-3
ç (DRN 01-415)
Figure 1.2-13 has been incorporated by reference in accordance with NEI 98-03.
Figure information can be found in Drawing G139.
Ï (DRN 01-415)
Revision 11 (05/01) WSES-FSAR-UNIT-3
ç (DRN 01-415)
Figure 1.2-14 has been incorporated by reference in accordance with NEI 98-03.
Figure information can be found in Drawing G140.
Ï (DRN 01-415)
Revision 11 (05/01) WSES-FSAR-UNIT-3
ç (DRN 01-415)
Figure 1.2-15 has been incorporated by reference in accordance with NEI 98-03.
Figure information can be found in Drawing G141.
Ï (DRN 01-415)
Revision 11 (05/01) WSES-FSAR-UNIT-3
ç (DRN 01-415)
Figure 1.2-16 has been incorporated by reference in accordance with NEI 98-03.
Figure information can be found in Drawing G142.
Ï (DRN 01-415)
Revision 11 (05/01) WSES-FSAR-UNIT-3
(EC-27161, R307)
Figure 1.2-17 has been incorporated by reference in accordance with NEI 98-03
Figure information can be found in Drawing G143
(EC-27161, R307)
Revision 307 (07/13)
WSES-FSAR-UNIT-3
ç (DRN 01-415)
Figure 1.2-19 has been incorporated by reference in accordance with NEI 98-03.
Figure information can be found in Drawing G145.
Ï (DRN 01-415)
Revision 11 (05/01) WSES-FSAR-UNIT-3
(EC-27161, R307)
Figure 1.2-20 has been incorporated by reference in accordance with NEI 98-03
Figure information can be found in Drawing G146
(EC-27161, R307)
Revision 307 (07/13)
WSES-FSAR-UNIT-3
ç (DRN 01-415)
Figure 1.2-23 has been incorporated by reference in accordance with NEI 98-03.
Figure information can be found in Drawing G149.
Ï (DRN 01-415)
Revision 11 (05/01) WSES-FSAR-UNIT-3
→ (DRC 01-415; LBDCR 16-043, R311)
Figure 1.2-24 has been incorporated by reference in accordance with NEI 98-03
Figure information can be found in Drawing G210
← ((DRC 01-415; LBDCR 16-043, R311)
Revision 311 (9/19)
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1.3 COMPARISONS
1.3.1 COMPARISONS WITH SIMILAR FACILITY DESIGNS
Table 1.3.1 presents a summary of the characteristics of Waterford 3 for Cycle 1. The table presents comparative data for San Onofre Units 2 and 3; Arkansas Nuclear One, Unit 2; and St. Lucie Unit 1.
The San Onofre Units 2 and 3, and Arkansas Nuclear One, Unit 2 designs were selected for comparison because of the basic similarity of the reactor cores and the Reactor Coolant Systems. Also they are well advanced in terms of licensing relative to Waterford 3. St. Lucie Unit 1 was selected because of the basic similarity in the containment design.
1.3.2 COMPARISON OF FINAL AND PRELIMINARY INFORMATION
1.3.2.1 General
This section contains a discussion of all significant changes that have been made in the Waterford 3 design since submittal of the PSAR until the docketing of the FSAR. Changes considered as significant include changes in design bases or criteria for safety-related structures, systems or components, plant arrangement, mode of system operation, type of equipment, or gross changes in component or system capacity. In general changes have been made to further increase the safety margin and operating flexibility of Waterford 3.
1.3.2.2 Site Characteristics
Additional site studies and field tests have been made since the submittal of the PSAR. No significant site characteristic changes have been brought to light that would require a design change.
1.3.2.3 Design Criteria
1.3.2.3.1 NRC Regulatory Guides
Many design changes are the result of the evolution of NRC's interpretation of the safety requirements to comply with the General Design Criteria of 10CFR50, Appendix A. This evolution has resulted in the promulgation of many Regulatory Guides which were not addressed in the PSAR. The intent of these new guides have been evaluated against the plant design and changes have been initiated where practical to bring Waterford 3 into general compliance.
1.3.2.3.2 Design Codes a) Containment Vessel Code Case (DRN 01-758, R11-A) The steel containment vessel has been designed to ASME Section III, Class "MC" rather than Class "B" requirements as stated in the PSAR. Class "B" was an earlier version of the ASME Code for containment vessels which was superseded by Class "MC" in 1971. The containment vessel design is described in Section 3.8. (DRN 01-758, R11-A) b) Changes in Concrete Quality Control Program
Concrete slump test frequency and maximum temperature have been changed in accordance with ASTM C-143 and ACI-305-72, respectively. The code requirements for concrete are discussed in Section 3.8
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1.3.2.3.3 Pipe Break Criteria
High and moderate energy piping systems have been analyzed in accordance with NRC Branch Technical Positions APCSB 3-1 and MEB 3-1 as discussed in LP&L letter LPL3690 dated July 11, 1975. (See Section 3.6.)
1.3.2.3.4 Tornado Design Criteria
As discussed in Section 3.3, protection against multiple tornado generated missiles has been extended to systems and components required for safe shutdown.
1.3.2.4 Reactor
Changes to fuel rod design have been made in order to reduce fuel densification problems. The number and design of fuel spacer grids has been changed to enhance fuel performances under seismic loading. Fuel design is described in Section 4.2.
1.3.2.5 Reactor Structures
1.3.2.5.1 Reactor Vessel Grillage Foundation
The reactor vessel grillage has been modified to incorporate a positive system of restraints to prevent upward motion of the reactor vessel following a LOCA. This Support System is discussed in Section 5.4.
1.3.2.5.2 Reactor Vessel Cavity
The design of the reactor vessel cavity has been modified to provide a reduction in neutron streaming from the cavity, and activation of Ar-40 in the containment atmosphere. The reactor vessel cavity is discussed in Sections 3.8 and 6.2.
1.3.2.6 Engineered Safety Features
1.3.2.6.1 Hot Leg Injection Capability
Provision has been made in the Safety Injection System to permit introduction of safety injection fluid during recirculation through both the hot and cold legs, thereby ensuring adequate long term post-LOCA cooling. The Safety Injection System is described in Section 6.3.
1.3-2
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1.3.2.6.2 Containment Ductwork Declassification
The containment ductwork ring header has been changed from Safety Class 2, seismic Category I to NNS, non-seismic Category I. The ring header is not required for the Containment Cooling System to perform its design function.
This duct system is discussed in Section 6.2.
1.3.2.6.3 Containment Vessel Installation, Testing and Inspection
Several changes have been made to the installation and testing procedures for the containment vessel. These changes include upgrading of containment vessel inner surface coating and cleaning requirements, deletion of radiography of randomly selected welds (15 percent) following postweld heat treatment, revision to welder qualification test record retention procedures, and acceptance of gas metal arc welding processes. The containment vessel is discussed in Section 6.2 and 3.8.
1.3.2.7 Instrumentation and Control
1.3.2.7.1 Reactor Protection System (RPS)
The Reactor Protection System (RPS) described in the PSAR has been expanded and some portions modified in order to provide automatic protection against axial xenon oscillations and to implement design improvements. a) The following changes were made to meet the requirement for automatic protection against axial xenon oscillations:
1) The high local power density trip is added;
2) The thermal margin/low pressure trip is replaced by the low DNBR trip;
3) The core protection calculators (CPCS) are added to provide the high local power density and low DNBR trips and the thermal margin/low pressure calculator is eliminated. b) As a consequence of the above addition of the CPCS, the following design changes are implemented:
1) The low reactor coolant flow trip function is incorporated into the low DNBR trip;
2) Reactor coolant flowrate is calculated by use of reactor coolant pump speed instead of being inferred by differential pressure measurements;
3) CEA position signals are incorporated into the RPS. c) A high logarithimic power level trip has replaced the high rate of change of power trip in order to provide improved protection against inadvertent boron dilution. The RPS also addresses the unplanned withdrawal of CEAs as the previous trip did.
1.3-3
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The Engineered Safety Feature Actuation System (ESFAS) has been changed in the following areas: a) The Emergency Feedwater Actuation Signal (EFAS) is added to the ESFAS. b) Diverse signals for Containment Isolation have been added. c) Variable setpoints for SIAS on low pressurizer pressure and Main Steam Isolation Signal on low steam generator pressure are added. d) The group testing capability is added.
Further discussion of the RPS is found in Section 7.2.
1.3.2.7.2 Core Operation Limit Supervisory System A non-safety-related Core Operating Limit Supervisory System (COLSS) has been added. The COLSS consists of sensors, algorithms implemented in the plant monitoring computer, and other equipment to monitor selected Nuclear Steam Supply System parameters and process the parameter information so that a comprehensive, on-line calculation of the margin to specified limiting conditions of operation is available at all times. The COLSS also provides the operator with an alarm so that he can maintain the reactor core within the limiting conditions of operation during steady-state operation by initiating a power reduction whenever any one of the monitored core conditions reaches its specified limiting condition of operation. This system is described further in Section 7.7. 1.3.2.7.3 Movable In-Core Instrument System
This system has been deleted.
1.3.2.7.4 Deletion of Containment Purge Isolation Signal (CPIS) and High Containments Radiation Input to the Plant Protection System
Since a fuel handling accident can occur only during shutdown, CPIS on high radiation should be independent of the Plant Protection System (PPS). Therefore, the PPS has been changed accordingly. This change also allows the deenergizing of the PPS for maintenance and inspection during shutdown. The PPS is discussed in Section 7.3.
1.3.2.8 Electric Power
Extensive redesign of all electrical and I&C cable trays, control boards and cabinets to meet the separation criteria of Regulatory Guide 1.75 (1/75) has been made. Compliance with RG 1.75 is discussed in Section 8.3.
1.3.2.9 Auxiliary Systems
1.3.2.9.1 High Density Spent Fuel Storage
Additional storage capacity of spent fuel has been provided for by means of high density poison racks. The design of these racks is described in Section 9.1.
1.3-4 Revision 9 (12/97)
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1.3.2.9.2 Fire Protection System
-012, R310) The Fire Protection System has been modified as a result of the new NRC requirements (10CFR50.48(c)). The Fire Protection System is discussed in Section 9.5. -012, R310) 1.3.2.10 Steam and Power Conversion System
There are no significant changes in the final design of the Steam and Power Conversion System from that described in the PSAR other than all volatile water treatment has been provided.
1.3.2.11 Radioactive Waste Management
Additional radwaste handling capability has been provided for by installing means to allow the use of portable solidification and demineralization systems.
1.3.2.12 Radiation Protection
Since the PSAR, the normal sampling system has been extensively rerouted and a new sampling panel with a modified equipment configuration has been introduced in the RAB design. A new sampling system has been installed for Post-Accident Sampling (P.A.S.S.) and additional shielding and sample tubing have been installed for this purpose. These systems are discussed in Subsection 12.5.3.
1.3.2.13 Conduct of Operations
Since the PSAR, extensive modifications to the Plant Security System have been made including the addition of much more sophisticated equipment, such as TV monitors and electronic card readers. The Plant Security System is summarized in Section 13.6.
A separate Administration Building has been added due to increases in the operating staff.
1.3.2.14 Initial Test and Operations
There are no significant changes in initial tests and operations affecting plant design from that described in the PSAR.
1.3.2.15 Accident Analyses
The methods used to analyze some of the accidents have been revised to take into account expansion of the RPS (see Subsection 1.3.2.7.1). In addition, there have been extensive refinements to computer codes, analytical investigations and tests to demonstrate compliance with 1OCFR50, Appendix K,. Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors.
The offsite accident doses presented in the PSAR were calculated using the atmospheric dispersion factors (X/Q) based on preliminary meteorological data. The dose analyses presented in Chapter 15 are based on X/Q values obtained from the onsite meteorological monitoring program described in Section 2.3. The onsite data demonstrates the conservatism of the X/Q values used in the PSAR accident analyses.
1.3.2.16 Technical Specifications
Details of safety limiting settings and limiting conditions of operation have changed as a result of changes enumerated in other parts of this subsection.
1.3-5 Revision 310 (12/17)
WSES-FSAR-UNIT-3
Subject coverage is in accordance with NRC Standard Technical Specifications as revised for Waterford 3.
1.3.2.17 Quality Assurance
The major change in the QA program since the submittal of the PSAR has been the commitment of WASH 1309, "Guidance on Quality Assurance Requirements During the Construction Phase of Nuclear Power Plants." Quality Assurance during operation is described in detail in the QA Program Manual.
1.3-6
WSES-FSAR-UNIT-3
TABLE 1.3-1 (Sheet 1 of 13) Revision 11-A (02/02) ¨(DRN 01 758) PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1 õ(DRN 01 758) Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1
Hydraulic and Thermal Design Parameters
Rated core heat output, Mwt 3,390 4.4 3,390 2,815 2,560
Rated core heat output, Btu/hr 11,570 x 106 4.4 11,570 x 106 9,608 X 106 8,737 X 106
Heat generated in fuel, % 97.5 4.4 97.5 96.5 97.5
System pressure, nominal, psia 2,250 4.4 2,250 2,250 2,250
System pressure, minimum steady state, psia 2,200 4.4 2,200 2,200 2,200
Hot channel factors,
Heat flux, Fq 2.35 2.35 2.35 2.85
Enthalpy rise, FH 1.55 4.4 1.55 1.55 2.02 DNB ratio at nominal conditions 2.07 (CE-1) 4.4 2.07 (CE-1) 2.26 (W-3) 2.30 (W-3)
Coolant flow Total flowrate, lb/hr 148 x 106 4.4 148 x 106 120.4 x 106 122 x 106
Effective flowrate for heat transfer, lb/hr 144.2 x 106 4.4 144.2 x 106 116.2 x 106 117.5 x 106
Effective flow area for heat transfer, ft2 54.7 4.4 54.7 44.7 53.5
Average velocity along fuel rods, ft/sec 16.4 4.4 16.4 16.4 13.6
Average mass velocity, lb/hr-ft2 2.64 x 106 4.4 2.64 x 106 2.60 x 106 2.20 x 106
Coolant temperatures, F Nominal inlet 553 4.4 553 553.5 538.9
Design inlet 553 4.4 556 556 5 544
Average rise in vessel 58 4.4 58 58.5 55
Average rise in core 60 4.4 60 60.5 56
Average in core 583 4.4 583 583.75 572 WSES-FSAR-UNIT-3
TABLE 1.3-1 (Sheet 2 of 13) Revision 9 (12/97)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1
Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1
Hydraulic and Thermal Design Parameters (Cont.)
Average in vessel 582 4.4 582 582.75 571.5
Nominal outlet of hot channel 642 4.4 642 652 640
Average film coefficient, Btu/hr-ft2-F 6,160 4.4 6,160 6,170 5,300
Average film temperature difference, F 30 4.4 30 31 35
Heat transfer at 100% power
Active heat transfer surface area, ft2 62,000 4.4 62,000 51,000 48,400
Average heat flux, Btu/hr-ft2 182,400 4.4 182,400 182,200 176,000
Maximum heat flux, Btu/hr/ft2 428,000 4.4 428,000 425,800 501,300
Average thermal output, KW/ft (Fuel Rod Only) 5.34 4.4 5.34 5.34 5.94
Maximum thermal output, KW/ft (Fuel Rod Only) 12.5 4.4 12.5 12.5 17
Maximum clad surface temperature at nominal 657.0 4.4 657.0 657 657 Pressure, F
Fuel center temperature, F maximum at 100% power 3,420 4.4 3,420 3,420 3,890
Core Mechanical Design Parameters
Fuel assemblies
Design CEA 4.2 CEA CEA CEA
Rod pitch, in. 0,506 4.2 0.5063 0.5063 0.58
Cross-section dimensions, in. 7.972 x 7.972 4.2 7.972 x 7.972 7.98 x 7.98 7.98 x 7.98 3 3 Fuel weight (as UO2 ), lbm 223.9 x 10 4.2 223.9 x 10 183,640 207,200
Total weight, lbm 310,744 4.2 314,867 256,827 271,280 Number of grids per assembly 11 4.2 11 12 8 WSES-FSAR-UNIT-3
TABLE 1.3-1 (Sheet 3 of 13) Revision 11-A (02/02)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1
Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1
Core Mechanical Desiltn Parameters (Cont.)
Fuel rods
Number 49,580 4.2 49,500 40,716 36,896 ¨(DRN 01 758) Outside diameter, in. 0.382 4.2 0,382 0,382 0.44
Diametral gap, in. 0.007 4.2 0,007 0,007 0.0085
Clad thickness, in. 0.025 4.2 0,025 0,025 0,026 õ(DRN 01 758)
Clad material Zircaloy-4 4.2 Zircaloy-4 Zircaloy Zircaloy
Fuel pellets
Material UO2 sintered 4.2 UO2 sintered UO2 sintered UO2 sintered Diameter, in. 0.325 4.2 0.325 0.325 0.3795
Length, in. 0.390 4.2 0.390 0.390 0.650
Control assemblies
Neutron absorber (See Table 4.2-1) 4.2 (See Table 4.2-1) B4 C/Ag-In-Cd B4 C/SS Cladding material Inconel 625 4.2 Inconel 625 NiCrFe alloy NiCrFe alloy
Clad thickness 0.035 4.2 0.035 0.035 0.040
Number of assembly, full/part-length 83/8 4.2 83/8 73/8 73/8
Number of rods per assembly 4,5/5 4.2 4,5/5 5 5
Nuclear Design Data
Structural characteristics
Core diameter, in. (equivalent) 136 4.2 136 123 136
Core height, in. (active fuel) 150 4.2 150 150 136.7 WSES-FSAR-UNIT-3
TABLE 1.3-1 (Sheet 4 of 13) Revision 9 (12/97)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1
Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1
Nuclear Design Data (Cont.)
H2 O/U, Unit cell (cold) 3.35 4.3 3.35 1.63 Number of fuel assemblies 217 4.2 217 177 217
UO2 Rods per assembly, unshimed/shimed Batch A 236 4.3 236 176 176
Batch B 236/220 4.3 236/220 164 164
Batch C 236/224 or 220 4.3 236/224 or 220 176/164 176/164/164
Performance characteristics loading technique 3-batch mixed 4.3 3-batch mixed 3-batch mixed 3-batch mixed central zone central zone central zone central zone
Fuel discharge burnup, MWD/MTU Average first cycle 12,731 4.3 12,731 12,500 12,800
Feed enrichment, wt% Region 1 1.87 4.3 1.87 1.93 1.93
Region 2 2.38 4.3 2.38 2.27 2.33
Region 3 2.88 4.3 2.88 2.94 2.82
Control characteristics effective multiplication (beginning of life)
Cold, no power, clean 1,170 4.3 1,170 1,182 1.170
Hot, no power, clean 1,125 4.3 1,125 1,136 1.134
Hot, full power, Xe equilibrium 1,067 4.3 1,067 1,075 1.078
Control assemblies Total rod worth (hot), % 11.35 4.3 11.35 12.3 11.0 WSES-FSAR-UNIT-3
TABLE 1.3-1 (Sheet 5 of 13) Revision 9 (12/97)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1
Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1
Boron concentrations for criticality:
Zero power no rods inserted, clean, ppm 832 832
Cold/Hot 899/832 4.3 899/832 1,004/987 945/935 At power with no rods inserted, 719/452 4.3 719/452 870/612 820/590
clean/equilibrium xenon, PPM
Kinetic characteristics, range over life
Moderator temperature coefficient, p/F See Table 4.3-4 4.3 See Table 4.3-4 -0.5 x 10-4 0.4 x 104 to to -3.1 x 10-4 -2.1 x 10-4
Moderator pressure coefficient, p/psi +0.7 x 10-6 4.3 +0.7 x 10-6 +0.45 x 10-6 +0.49 x 10-6 to to +2.97 x 10-6 +2.55 x 10-6
Moderator void coefficient, p/% Void -0.36 x 10-3 4.3 -0.36 x 10-3 -0.28 x 10-3 -0.26 x 10-3 to to -1.47 x 10-3 -1.35 x 10-3
Doppler coefficient, p/F -1.13 x 10-5 4.3 -1.13 x 10-5 -1.18 x 10-5 -1.45 x 105 to to to to -1.67 x 10-5 1.67 x 10-5 -1.78 x 10-5 -1.07 x 10-5
Reactor Coolant System-Code Requirements
Component
Reactor vessel ASME III Class 1 5.2 ASME III Class 1 ASME III Class A ASME III Class A
Steam generator
Tube side ASME III Class 1 5.2 ASME III Class 1 ASME III Class A ASME III Class A
Shell side ASME III Class 2 5.2 ASME III Class 2 ASME III Class A ASME III Class A WSES-FSAR-UNIT-3
TABLE 1.3-1 (Sheet 6 of 13) Revision 9 (12/97)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1
Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1
Nuclear Design Data (Cont.)
Pressurizer ASME III Class 1 5.2 ASME III Class 1 ASME III Class 1 ASME III Class A
Pressurizer relief (or quench) tank ASME VIII Div. 1 5.4 ASME VIII Div. 1 ASME III Class C ASME III Class C
Pressurizer safety valves ASME III Class 1 5.2 ASME III Class 1 ASME III Class A ASME III Class
Reactor coolant piping ASME III Class 1 5.2 ASME III Class 1 ASME III Class 1 USAS B31.7 (USAS B31.1) (USAS B31.1) Principal Design Parameters of the Reactor Coolant System
Operating pressure, psig 2,235 5.1 2,235 2,235 2,235
Reactor inlet temperature, F 553 5.1 553 553.5 539.7
Reactor outlet temperature, F 611.2 5.1 611.2 612.5 595.1
Number of loops 2 5.1 2 2 2
Design pressure, psig 2,485 5.1 2,485 2,485 2,485
Design temperature, F 650 5.1 650 650 650
Hydrostatic test pressure (cold), psig 3,110 3,110 3,110 3,110
Total coolant volume, ft3 10,300 (without 10,300 (without 9,376 11.101 pressurizer) pressurizer) Principal Design Parameters of the Reactor Vessel
Material See Table 5.2-3 5.2 See Table 5.2-2 SA-533, Grade B SA-533, Grade B Class 1, low Class, 1, low alloy steel, alloy steel, internally clad internally clad with Type 304 with Type 304 austenitic SS austenitic SS WSES-FSAR-UNIT-3
TABLE 1.3-1 (Sheet 7 of 13) Revision 9 (12/97)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1
Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1 Principal Design Parameters of the Reactor Vessel (cont’d)
Design pressure, psig 2,485 5.4 2,485 2,485 2,485
Design temperature, F 650 5.4 650 650 650
Operating pressure, psig 2,235 5.4 2,235 2,235 2,235
Inside diameter of shell, in. 172 5.4 172 157 172
Outside diameter across nozzles, in. 253 253 238 253
Overall height of vessel and enclosure head, 43-6-1/2 5.4 43-6-1/2 43-4-1/6 41-11-3/4 ft-in. to top of CEDM nozzle
Minimum clad thickness, in. 1/8 5.4 1/8 1/8 5/16
Principal Design Parameters of the Steam Generators Number of Units 2 5.4 2 2 2
Type Vertical U-tube 5.5 Vertical Untube Vertical U-tube Vertical U-tube with integral with integral with integral with integral moisture separator moisture separator moisture seperator moisture separator
Tube material Inconel(ASME SB-163) 5.4 Inconel(ASME SB-163) NiCrFe alloy NiCrFe alloy
Shell material SA-533 Gr. B SA-533 Gr B SA-533 Gr B SA-533 Gr. B Class 1 and Class 1 and Class 1 and Class 1 and SA-516, Gr. 70 SA-516, Gr. 70 SA-516, Gr. 70 SA-516, Cr. 70
Tube side design pressure, psig 2,485 5.4 2,485 2,485 2,485
Tube side design temperature, F 650 5.4 650 650 650
Tube side design flow, lb/hr 74 x 106 5.4 74 x 106 60.2 x 106 61 x 106
Shell side design pressure, psia 1,100 5.4 1,100 1,100 1,000
Shell side design temperature, F 560 5.4 560 560 550 WSES-FSAR-UNIT-3
TABLE 1.3-1 (Sheet 8 of 13) Revision 9 (12/97)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1
Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1
Principal Design Parameters of the Steam Generators (Cont)
Operating pressure, tube side, nominal, psig 2,235 5.4 2,235 2,235 2,235
Operating pressure, shell side, maxim- , psig 985 985 985 885
Maximum moisture at outlet at full load, % 0.2 5.4 0.2 0.2 0.2
Hydrostatic test pressure, tube side (cold) psig 3,110 3,110 3,110 3,110
Steam pressure, at full power, psia 900 5.4 900 900 815
Steam temperature, at full power, F 532 5.4 532 531.95 520.3
Principal Design Parameters of the Reactor Coolant Pumps
Number of units 4 5.4 4 4 4
Type Vertical, single Vertical, single Vertical, single Vertical, single stage radial flow stage radial flow stage centrifugal stage centrifugal with bottom with bottom with bottom with bottom suction and suction and suction and suction and horizontal horizontal horizontal horizontal discharge discharge discharge discharge
Design pressure, psig 2,485 5.4 2,485 2,45 2,485
Design temperature, F 650 5.4 650 650 650
Operating pressure, nominal psig 2,235 5.4 2,235 2,235 2,235
Suction temperature, F 553 5.4 553 553.5 540
Design capacity, gal/min 99,000 5.4 99,000 80,000 80,000
Design head, ft 310 5.4 310 275 250
Hydrostatic test Pressure (cold), psig 3,110 3,110 3,110 3,110 WSES-FSAR-UNIT-3
TABLE 1.3-1 (Sheet 9 of 13) Revision 9 (12/97)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1
Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1
Principal Design Parameters of the Reactor Coolant Pumps (Cont’d)
Motor type AC induction, AC induction, AC induction, AC induction, single speed single speed single speed single speed
Motor rating, hp 9,700 9,700 6,500 6,500
Principal Design Parameters of the Reactor Coolant Piping
Material SA-516, Gr 70 SA-516, Gr 70 SA-516, Gr 70 with nominal with nominal with nominal 7/32 SS Clad 7/32 SS Clad 7/32 SS Clad
Hot leg ID, in. 42 5.4 42 42 42
Cold leg ID, in. 30 5.4 30 30 30
Between pump and steam generator ID, in. 30 5.4 30 30 30
Engineered Safety Feature
Safety injection system
No. of high pressure pumps 3 6.3 3 3 3
No. of low pressure pumps 2 6.3 2 2 2
Containment spray
No. of pumps 2 6.2 2 2 2
Containment fan coolers No. of units 4 6.2 4 4 4
Air flow capacity, each at emergency conditions, ft3/min 35,000 6.2 31,000 50,000 55,800
Safety injection tanks, number 4 6.3 4 4 4
Emergency power Diesel-generator unit 2 8.3 4 (for two units) 2 2 WSES-FSAR-UNIT-3
TABLE 1.3-1 (Sheet 10 of 13) Revision 9 (12/97)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1
Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1
Principal Design Parameters of the Reactor Coolant Piping (Cont’d)
Containment System Parameters Type Steel containment Steel-lined Steel-lined Steel containment vessel with prestressed post prestressed post vessel with cylindrical cylindrical shell, tensioned con- tensioned con- shell, hemisperical hemisperical dome crete cylinder, crete cylinder, dome and ellipsoidal and ellipsoidal curve dome roof. curved dome roof bottom - ASME Code, bottom - ASME Code, Section III, Class B, Section III, Class MC, surrounded by reinforced surrounded by concrete Shield Building reinforced concrete Shield Building
Design Parameters - Containment
Inside Diameter, ft. 140 3.8 150 116 140
Height, ft. 240.5 3.8 172 207 232
Free volume, ft3 2,677,000 6.2 2,335,000 1,780,000 2,500,000
Reference accident Pressure, psig 44 3.8 60 54 44
Steel Thickness, in. - - - - Vertical Wall 1.90 3.8 Not applicable Not applicable 1.91 Hemispherical Head 0.95 Not applicable Not applicable 0.95 Knuckles 2.25 Not applicable Not applicable 2.25
Concrete Thickness, ft. - - - Vertical Wall Not Applicable 3.8 4 1/3 3 3/4 Not applicable
Dome Not Applicable 3 3/4 3 1/4 Not applicable WSES-FSAR-UNIT-3
TABLE 1.3-1 (Sheet 11 of 13) Revision 9 (12/97)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1
Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1
Principal Design Parameters of the Reactor Coolant Piping (Cont’d)
Design Parameters - Shield Building 3.8 Not applicable Not applicable
Inside Diameter, ft- 148 148
Height, ft. (top of foundation to top of dome) 249.5 230.5
Concrete Thickness, ft. Vertical Wall 3 3 Dome 2.5 2.5
Containment Leak Prevention and Mitigation Systems Leak-tight pene- 6.2 Leak-tight pen- Leak-tight pene- Leak-tight pene- tration Automatic tration, and tration, and tration. Automatic isolation where continuous steel continuous stee isolation where required. liner. Automatic liner. Automatic required. isolation where isolation where required. The ex- required. haust from pene- tration rooms to vent.
Gaseous Effluent Purge Discharge through 6.2 Discharge thru Discharge thru Discharge thru vent. vent. vent. vent.
RADIOACTIVE WASTE MANAGEMENT SYSTEM Liquid Waste Processing Systems Reactor Coolant Waste Holdup Tank 11.2 (BMS) Number 4 1/2 4 4
Capacity (Gal.), each 45,000 6,000/25,000 51,270 40,000 WSES-FSAR-UNIT-3
TABLE 1.3-1 (Sheet 12 of 13) Revision 11-B (06/02)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1
Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1
RADIOACTIVE WASTE MANAGEMENT SYSTEM (Cont'd)
Degasifier (DRN 00 803) Number Flash Tank* 1 (Gas Stripper) 1 Flash Tank ê(DRN 00 803) Capacity (gpm) - - Concentrators Number 2 1 (For 2 units) 1 1 Capacity (gpm) 20 50 gpm 20 2
Gaseous Waste Processing Systems Waste Gas Decay Tank 11.3 Number 3 3 6 (For 2 units) 3 3 Capacity (ft ), each 600 500 300 144 Pressure (psig) 380 150 380 190 Hold-up Time (days) 60 30 30 30
ELECTRIC SYSTEMS
Number of Offsite Circuits 7 8.2.1.1 8 3 3
Number of Incoming Lines to Startup Transformers 2 8.2 2 2 2
Number of Startup Transformers 2 8.2 4 1+1(shared) 2
Number of Main Unit Transformers (Three Phase) 2 8.2 1 3 (single phase) 2
Number of 4.16 KV Engineered Safety Features System Buses 3 8.3 3 2 3
Number of 480V Engineered Safety Features System Buses (Power Centers) 3 8.3 3 2 4
Number of 120V AC Vital Buses 8 8.3 4 4 3
Number of Standby Diesel Generators 2 8.3 2 2 2
Diesel Generator Rating (KW) 4400 8.3 4700 2850 3500
(DRN 00 803) (*) The Flash Tank is inactive per ER-W3-00-0225-00-00. ê(DRN 00 803) WSES-FSAR-UNIT-3
TABLE 1.3-1 (Sheet 13 of 13) Revision 11-A (02/02)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1
Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1
INSTRUMENTATION SYSTEMS*
Reactor Protective System 7.2 7.2 7.2 7.2
Reactor and Reactor Coolant System 7.7.1.1 7.7.1.1 7.7.1.1 7.7.1.1 7.1.1.2 7.7.1.2 7.7.1.2 7.7.1.2
Steam and Feedwater Control System 7.7.1.3 7.7.1.3 7.7.1.3 7.7.1.3
Nuclear Instrumentation 7.2.1.1 7.2.1.1 7.2.1.1 7.2.1.1
Non-Nuclear Process Instrumentation 7.5.1.5 7.5.1.5 7.5.1.5 7.5.1.5
CEA Position Instrumentation 7.5.1.3 7.5.1.3 7.5.1.3 7.5.1.3
¨ (DRN 01 758) * This section is not suited for tabular description. SAR section numbers have been included for the location of the detailed description of each system. õ (DRN 01 758) WSES-FSAR-UNIT-3
1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS
Louisiana Power & Light Company, as owner, has arranged for the purchase of equipment and consulting, engineering, and construction services for the installation of Waterford Steam Electric Station Unit No. 3. As owner, Louisiana Power & Light Company is responsible for the design, construction, and operation of the unit.
Ebasco Services Incorporated has been retained for engineering, procurement and management of construction services. Ebasco is also providing assistance in obtaining licenses and permits, in preoperational testing, in quality control, and in initial startup of the plant. (DRN 01-758) Combustion Engineering, Inc. has been contracted to design, manufacture, and deliver to the site a complete Nuclear Steam Supply System including first core fuel. In addition, Combustion Engineering, Inc. is supplying competent technical and professional consultation for erection, initial fuel loading, testing and initial startup of the complete Nuclear Steam Supply System. Combustion Engineering, Inc. is participating in plant personnel training. (DRN 01-758) The Westinghouse Electric Corporation is supplying the turbine generator and its auxiliaries.
1.4-1 Revision 11-A (02/02)
WSES-FSAR-UNIT-3 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION
This section provides a description of safety related technical information relevant to this application. Combustion Engineering, Inc., (C-E), is conducting research and development programs relating to the requirements of this section.
Waterford 3 reactor incorporates a 16 x 16 fuel assembly design with five guide tubes. This design provides an increase in conservatism for loss of-coolant accident (LOCA) considerations with a minimum change from previous C-E fuel designs. Previous designs have undergone extensive testing, and operating experience is now being acquired.
The three test programs described in Subsections 1.5.1, 1.5.2, and 1.5.3 are considered necessary to confirm the adequacy of the 16 x 16 fuel assembly design. (DRN 01-758, R11-A) CENPD-84(1), CENPD-143(2), CENPD-184(3), CENPD-299(4) present descriptions of development programs aimed at verifying the Nuclear Steam Supply System (NSSS) design and the anticipated performance characteristics and at confirming the design margins. Other programs that apply to this plant are identified in Subsection 1.5.4 through 1.5.8. (DRN 01-758, R11-A) 1.5.1 FRETTING AND VIBRATIONS TESTS OF FUEL ASSEMBLIES
Extensive autoclave vibration and dynamic flow tests have been performed to characterize fuel rod and spacer grid fretting corrosion in C-E fuel assemblies. The results of these tests are discussed in more detail in Subsection 4.2.3.2.4.
Tests have been completed using a full sized 16 x 16 fuel assembly. This assembly is similar to the 16 x 16 five guide tube design used on the Waterford 3 reactor. This assembly was subjected to flow testing under conditions of temperature, water chemistry, pressure, and flow velocities in excess of normal reactor conditions. Further information is provided in Subsections 4.2.3.1.1, 4.2.3.1.2, and 4.2.4.4.
1.5.2 DEPARTURE FROM NUCLEATE BOILING (DNB) TESTING (DRN 01-758, R11-A; EC-13881, R304) Extensive heat transfer testing has been completed with electrically heated rod bundles representative of the C-E 16 x 16 and 14 x 14 fuel assemblies. The program for each assembly geometry included tests to determine the effects on DNB of the control element assembly (CEA) guide tube, bundle heated length, and grid spacing, and lateral and axial power distributions. Each test yielded DNB data over a wide range of conditions of interest for pressurized water reactor (PWR) design. Those data were used with the TORC subchannel analysis code to develop and to verify the CE-1 DNB correlation for predicting DNB in fuel assemblies with standard spacer grids. The CE-1 correlation (for standard assemblies) and the ABB- NV correlation (for NGF assemblies), which are discussed in more detail in Subsection 4.4.4.1, are used in computing margin to DNB for Waterford 3. (DRN 01-758, R11-A; EC-13881, R304)
1.5.3 FUEL ASSEMBLY STRUCTURAL TESTS
The fuel assembly structural testing program was designed to verify the structural adequacy of the fuel assembly design under normal handling, normal operation, seismic excitation, and LOCA loadings. The test program provides the structural characteristics employed in the fuel assembly structural analyses.
1.5-1 Revision 304 (06/10)
WSES-FSAR-UNIT-3
A series of tests were conducted on a 14 x 14 fuel assembly to determine the combined axial and lateral load deflection characteristics of the fuel assembly. Axial compression tests and axial drop tests were performed. Measurements were made of axial loads, axial deflections, lateral deflections of all spacer grids, and strains in the guide tubes and fuel rods.
A series of structural tests on the 16 x 16 fuel assembly design was also conducted. The fuel assembly was subjected to both static and dynamic tests so as to determine basic structural characteristics. In addi- tion, several 16 x 16 spacer grids were subjected to impact tests to determine dynamic load deflection characteristics and damage limits. These tests are also discussed in Subsection 4.2.3.1.3.
1.5.4 FUEL ASSEMBLY FLOW MIXING TESTS
The objective of the fuel assembly flow mixing program was to obtain information on the magnitude of coolant mixing in C-E fuel assemblies. Several series of tests have been completed, and the data from these tests provide a sound basis for the treatment of coolant mixing in design thermal margin calculations.
The first series of single phase flow mixing tests was run in 1966 with a prototype C-E PWR fuel assembly. The average level of coolant mixing was determined using dye injection and sampling equipment.
A second series of single phase mixing tests was conducted in 1968 with a model representing a portion of a 14 x 14 CEA-type fuel assembly. Those tests, which also used dye injection and sampling techniques, are described in Reference 1.
More recently, tests were conducted in which coolant temperatures were measured in the subchannels of electrically heated rod bundles representative of the 14 x 14 or 16 x 16 fuel assemblies with standard spacer grids.
As discussed in Subsection 4.4.4.1, those data provide confirmation that the results from the previous dye sampling experiments are applicable for the fuel assembly design used in Waterford 3.
1.5.5 REACTOR FLOW MODEL TESTING AND EVALUATION
The objective of the reactor flow model test programs is to obtain information on: a) Flow and pressure distributions in various regions of the reactor b) Pressure loss coefficients c) Hydraulic loads on certain vessel internal components
This information is used for establishing or verifying design hydraulic parameters.
1.5-2 Revision 11-A (02/02)
WSES-FSAR-UNIT-3
Flow model testing, which began in 1966, was designed to obtain those reactor hydraulic design data not amenable to direct calculation. Scale model testing possesses the advantages, relative to actual reactor tests, of:
Providing the information early in the design stage
Being more suitable for extensive instrumentation
Being flexible so that proposed design modifications can be investigated
The reactor flow models used by C-E are generally 1/5 true scale models. In the first four C-E flow model programs, a closed-core design was used. The closed core simulates the reactor fuel assemblies with individual closed wall tubes containing orifices to provide the correct axial hydraulic resistance. Conclusions from the tests on the first four model configurations are summarized in Reference 1. A 1/5 scale flow model, representative of Waterford 3, was tested in 1976. This model has an open core design. Further discussion of the C-E flow model test programs is provided in Subsection 4.4.4.2.1.
1.5.6 FUEL ASSEMBLY FLOW TESTS
The objectives of the fuel assembly flow test program included assessment of the effect of postulated flow maldistributions on thermal behavior and margin.
The program originated in 1967 with fuel assembly flow distribution testing. Both flow visualization and flow pattern measurements were generated on an overscale model of the lower portion of an early C-E design fuel assembly.
A second test series was conducted for the CEA type fuel assembly. The second test series was designed to:
Determine the effect of flow obstructions on flow distribution within the fuel assembly
Determine the magnitude of the effect of the disturbed flow patterns on the thermal margin within a CEA type fuel assembly
The information from these tests, described further in Reference 1, has established the effect of flow obstructions within the fuel assembly. Additional information on the effects of postulated fuel coolant channel flow blockages is presented in Subsection 4.2.3.2.16.
1.5.7 CONTROL ELEMENT DRIVE MECHANISM (CEDM) TESTS
Performance testing of the magnetic jack CEDM is described in Subsections 3.9.4.4.1 and 4 2.4.4 and in Reference 1. The program has confirmed the operability of the drive assembly in normal and misaligned conditions as well as the load carrying capability and life characteristics.
1.5-3
WSES-FSAR-UNIT-3
1.5.8 DNB IMPROVEMENT
The DNB improvement program was initiated by C-E in order to obtain empirical information on the departure from nucleate boiling (DNB) phenomenon and on other thermal and hydraulic characteristics of C-E fuel assemblies. Testing has been performed with electrically heated rod bundles that correspond dimensionally to fuel rod configurations under in-reactor temperature pressure and flow conditions to obtain data on DNB, pressure drop, and coolant channel exit temperatures. These data were employed to verify that the C-E thermal hydraulic design methods conservatively predict DNB.
The DNB improvement program is described in References 1, 2, 3 and 4. It is a continuing program providing improvements in the accuracy of C-E thermal and hydraulic computer programs for predicting local coolant conditions and pressure drops and confirming the applicability of currently used DNB correlations to the C-E fuel design. Additional information on the program and results applicable to Waterford 3 are presented in Subsection 4.4.4.1.
SECTI0N 1.5: REFERENCES
1. "Safety Related Research and Development for Combustion Engineering Pressurized Water Reactors, Program Summaries," CENPD-87 and CENPD-87, Rev. 01 (Nonproprietary) (Proprietary) transmitted to DL by letter, Mr. F.M. Stern to Mr. R.C. DeYoung, Jr., March 18, 1973.
2. "Safety Related Research and Development for C-E Pressurized Water Reactors 1974 Program Summaries," Combustion Engineering Topical Report, CENPD-143 (Proprietary) and CENPD-143, Rev. 01, (Nonproprietary), May, 1974.
3. "Safety Related Research and Development for Combustion Engineering Pressurized Water Reactors - 1974 Program Summaries," CENPD-184-P (Proprietary) and CENPD-184 (Nonproprietary) May, 1975.
4. "Safety Related Research and Development for Combustion Engineering Pressurized Water Reactors, 1975 Program Summaries," CENPD-229-P (Proprietary) and CENPD-229 (Nonproprietary), June, 1976.
1.5-4
WSES-FSAR-UNIT-3
1.6 MATERIAL INCORPORATED BY REFERENCE
The following topical reports are incorporated by reference.
Report FSAR Number Author and Title Date to NRC Section
CENPD-26 Combustion Engineering, Inc. August 1971 3.6, 6.2 (with Suppl. "Description of Combustion 1 through 5) Engineering Loss of Coolant Calculational Procedures"
CENPD-42 Combustion Engineering, Inc. August 1972 3.6 "Dynamic Analysis of Reactor Vessel Internals Under Loss of Coolant Accident Conditions with Application to C-E 800 Mwe Class Reactors"
CENPD-67 Combustion Engineering, Inc. September 1973 10.3 (with Suppl. "Iodine Decontamination 1 and 2 and Factors During PWR Steam Addenda 1 Generation and Steam Venting" and 2)
CENPD-87 "Safety Related Research Development 1.5 for CE PWR's, Program Summaries"
CENPD-98 Combustion Engineering, Inc. July 1973 4.4, 15.0 "Coast Code Description"
CENPD-105 Combustion Engineering, Inc. June 1973 4.3 "Fast Neutron Attenuation by the ANISN-SHADRAC Analytical Method"
CENPD-107 Combustion Engineering, Inc. August 1974 15.0 (with Suppl. "CESEC" 1 through 4)
CENPD-118 Combustion Engineering, Inc. September 1974 4.3 "Densification of Combustion Engineering Fuel"
CENPD-132 Combustion Engineering, Inc. September 1974 6.2, 15.6 (with Suppl. "Calculative Methods for the 6.3 1 and 2) C-E Large Break LOCA Evaluation Model"
1.6-1 WSES-FSAR-UNIT-3
Report FSAR Number Author and Title Date to NRC Section
CENPD-133 Combustion Engineering, Inc. September 1974 6.2, 15.6 (with "CEFLASH-4A Fortran IV 6.3 Suppl. 2) Digital Computer Program for Reactor Blowdown Analysis"
CENPD-134 Combustion Engineering, Inc. September 1974 6.2, 15.6 (with "COMPERC-II A Program for 6.3 Suppl. 1) Emergency Refill-Reflood of the Core"
CENPD-135 Combustion Engineering, Inc. September 1974 4.2, 6.3, (with "STRIKIN-II A Cylindrical 15.6 Suppl. Geometry Fuel Rod Heat 2 and 4) Transfer Program"
CENPD-136 Combustion Engineering, Inc. September 1974 4.2, 15.6 "High Temperature Properties of Zircaloy and UO2, for use in LOCA Evaluation Model"
CENPD-137 Combustion Engineering, Inc. September 1974 6.3, 15.6 (with "Calculative Methods for the Suppl. 1) C-E Small Break LOCA Evaluation Model" (DRN 01-758) CENPD-138 "PARCH, A FORTAN IV Digital Program February 1975 15.6 to Evaluate Pool Boiling, Axial Rod and Coolant Heatup" (with Supplement 1) (DRN 01-758) CENPD-139 Combustion Engineering, Inc. September 1974 4.1, 4.2, (with "C-E Fuel Evaluation Model" 4.3, 6.3 Suppl. 1)
CENPD-145 Combustion Engineering, Inc. April 1975 4.3, 7.7 "A Method of Analyzing In-Core Detector Data in Power Reactors"
CENPD-148 Combustion Engineering, Inc. November 1974 4.6 "Review of Reactor Shutdown System (PPS Design) for Common Mode Failure Susceptibility"
CENPD-153 Combustion Engineering, Inc. August 1974 4.3 "Evaluation Uncertainty in the Nuclear Form Factor
1.6-2 Revision 11-A (02/02)
WSES-FSAR-UNIT-3
Report FSAR Number Author and Title Date to NRC Section
CENPD-153 Measured by Self Powered (con't) Fixed In-Core Detector Systems"
CENPD-155 Combustion Engineering, Inc. October 1974 5.3 "C-E Procedure for Design, Fabrication, Installation and Inspection of Surveillance Specimen Brackets Attached to Reactor Vessel Beltline Region"
CENPD-161 Combustion Engineering, Inc, June 1975 4.1, 4.2, "TORC - A Computer Code for 4.4, 15.0 Determining the Thermal Margin of a Reactor Core"
CENPD-162 Combustion Engineering, Inc. May 1975 4.4 (with "CHF Correlation for C-E Fuel Suppl. 1) Assemblies with Standard Spacer Grids - Part 1; Uniform Axial Power Distribution"
CENPD-168 Combustion Engineering, Inc. September 1976 3.6, 6.2 Rev. 1 "Design Basis Pipe Breaks for the Combustion Engineering Two Loop Reactor Coolant System"
CENPD-169 Combustion Engineering, Inc. August 1975 4.3, 7.2 "Assessment of the Accuracy 7.7 of PWR Operating Limits as Determined by Core Operating Limit Supervisory System"
CENPD-170 Combustion Engineering, Inc. August 1975 7.2 "Assessment of the Accuracy of the PWR Safety System Actuation as Performed by the Core Protection Calculators"
CENPD-178P Combustion Engineering, Inc. October 1976 3.9, 4.2 and 178 "Structural Analysis of the 16 x 16 Fuel Assembly for Combined Seismic and Loss- of-Coolant-Accident Loadings"
1.6-3
WSES-FSAR-UNIT-3
Report FSAR Number Author and Title Date to NRC Section
CENPD-179 Combustion Engineering, Inc. April 1976 4.2 "C-E Thermo-Structural Fuel Evaluation Method"
CENPD-183 Combustion Engineering, Inc. August 1975 15.0 "C-E Methods for Loss of 15.3 Flow Analysis"
CENPD-187 Combustion Engineering, Inc. October 1975 4.2 (with "Method of Analyzing Creep Suppl. 1) Collapse of Oval Cladding"
CENPD-190 Combustion Engineering, Inc. January 1976 15.4 "C-E Method for Control Element Assembly Ejection Analysis"
CENPD-198P Combustion Engineering, Inc. December 1975 4.2 and 198 "Zircaloy Growth-In-Reactor Dimensional Changes in Zircaloy-4 Fuel Assemblies"
CENPD-206 Combustion Engineering, Inc. December 1976 4.4 "Comparison of TORC Code Predictions with Experimental Data"
CENPD-207 Combustion Engineering, Inc. June 1976 4.4 "Critical Heat Flux Corre- lation for C-E Fuel Assemblies with Standard Spacer Grids, Part 2, Non-Uniform Axial Power Distributions"
CENPD-213 Combustion Engineering, Inc. February 1976 6.3, 15.6 "Application of FLECHT Reflood Heat Transfer Coefficients to Combustion Engineering 16 x 16 Fuel Bundles"
CENPD-225P Combustion Engineering, Inc. October 1976 4.2, 4.4 and 225 "Fuel and Poison Rod Bowing"
CENPD-252 "Method for the Analysis of Blowdown July 1979 3.9E Induced Forces in a Reactor Vessel"
1.6-4
WSES-FSAR-UNIT-3
Report FSAR Number Author and Title Date to NRC Section
WCAP 7709-L Electric Hydrogen Recombiners April 1972 6.2.5 for PWR Containments
(DRN 03-2054, R14) CEN-367-A Combustion Engineering, Inc. February 1991 3.6 Leak-Before-Break of Primary Coolant Loop Piping in Combustion Engineering Designed Nuclear Steam Supply System
(DRN 03-2054, R14) (EC-13881, R304) WCAP-11596-P-A Westinghouse Electric Company, June 1988 4.2, 4.3A, “Qualification of the PHOENIX – 15.1 P/ANC Nuclear Design System for Pressurized Water Reactor Cores”
WCAP-10965-P-A Westinghouse Electric Company, September 1986 4.2, 4.3A, “ANC: A Westinghouse Advanced Nodal 15.1 Computer Code”
WCAP-10965-P-A Westinghouse Electric Company, April 1989 4.2, 4.3A, Addendum 1 “ANC: A Westinghouse Advanced Nodal 15.1 Computer Code: Enhancements to ANC Rod Power Recovery”
WCAP-16045-P-A Westinghouse Electric Company, August 2004 4.3A “Qualification of the Two- Dimensional Transport Code PARAGON,”
WCAP-16072-P-A Westinghouse Electric Company, August 2004 4.2, 4.3A “Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs”
CENPD-404-P-A Combustion Engineering, Inc., November 2001 4.2, 4.3A, “Implementation of ZIRLO Material 15.0 Cladding in CE Nuclear Power Fuel Assembly Designs,”
WCAP-16500-P-A Westinghouse Electric Company, August 2007 4.2, 4.3A “CE 16 x 16 Next Generation Fuel Core Reference Report, “
WCAP-12610-P-A Westinghouse Electric Company, July 2006 4.2, 4.3A, and CENPD-404- “Optimized ZIRLOTM,” 15.0 P-A Addendum 1- A EC-13881, R304)
1.6-5 Revision 304 (06/10)
WSES-FSAR-UNIT-3
(EC-13881, R304) Report FSAR Number Author and Title Date to NRC Section
WCAP-16523-P- Westinghouse Electric Company, August 2007 4.4, 15.0, A “Westinghouse Correlations 15.1, 15.2, WSSV and WSSV-T for Predicting 15.3, 15.4 Critical Heat Flux in Rod Bundles with Side-Supported Mixing Vanes,”
CENPD-387-P-A Combustion Engineering, Inc., May 2000 4.3A, 15.0 “ABB Critical Heat Flux Correlations for PWR Fuel,”
WCAP-15996-P- Westinghouse Electric Company, March 2005 15.0, 15.1, A, Rev. 1 “Technical Manual for the CENTS 15.2, 15.3, Code,” 15.4
CEN-356(V)-P-A, Combustion Engineering, Inc., May 1988 4.2, 4.3A Revision 01-P-A “Modified Statistical Combination of Uncertainties, (EC-13881, R304) (EC-19087, R305) WCAP-17817-P Technical Justification for Eliminating February 2010 3.6.3 Pressurizer Surge Line Rupture as the Structural Design Basis for Waterford Steam Electric Station, Unit 3 Using Leak-Before- Break Methdology (EC-19087, R305)
1.6-6 Revision 305 (11/11)
WSES-FSAR-UNIT-3
(DRN 04-1444, R13-B) The following information is historical pursuant to NEI 98-03 and is identified by a designation of “start” and “end”.
Start of historical data. (DRN 04-1444, R13-B)
1.7 ELECTRICAL, INSTRUMENTATION, AND CONTROL DRAWINGS
Table 1.7-1 is a list of proprietary and nonproprietary electrical, instrumentation and control (EI&C) drawings. Three copies of all proprietary and seven copies of all nonproprietary drawings have been submitted under separate cover.
Drawing 1564-B-431 (Sheet 1 thru 11) listed in Table 1.7-1 provides symbols for flow diagrams and instrument schematics and logic diagrams.
(DRN 04-1444, R13-B) End of historical data. (DRN 04-1444, R13-B)
1.7-1 Revision 13-B (01/05)
G(DRN 02 85, R11 A; 04 1444, R13 B) The following table is historical pursuant to NEI 98-03 and is identified by a designation of “start” and “end”.
Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A)
TABLE 1.7-1 (Sheet 1 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (A) Instrument Location Arrangement Drawings
G-428 1 7 11-81 E Turbine Building 2 6 12-80 E Turbine Building 3 6 10-80 E Reactor Building 4 5 10-80 E Reactor Building 5 6 12-80 E Reactor Building 6 5 10-80 E Reactor Building 7 6 10-80 E Reactor Building 8 7 01-82 E Reactor Building 9 5 01-82 E Reactor Building 10 5 11-81 E Reactor Building
G-429 1 5 01-82 E Miscellaneous Area 2 1 12-80 E Miscellaneous Area 3 2 12-80 E Miscellaneous Area
G-432 1 7 11-81 E Reactor Auxiliary Building 2 5 12-80 E Reactor Auxiliary Building 3 6 05-81 E Reactor Auxiliary Building 4 5 12-80 E Reactor Auxiliary Building 5 6 12-80 E Reactor Auxiliary Building 6 6 07-78 E Reactor Auxiliary Building 7 6 01-82 E Reactor Auxiliary Building 8 7 01-82 E Reactor Auxiliary Building
C-433 5 06-78 E Fuel Handling Building
G-435 1 2 08-78 E No Sampling System Instr. Arrgt. (Turbine Building) 2 2 11-78 E No Sampling System Instr. Arrgt. (Reactor Aux. Building) 3 2 6-23-78 E Sampling System (Reactor Aux. Bldg.) 4 2 7-21-78 E Sampling System (Reactor Bldg.) 5 2 7-21-78 E Sampling System (Reactor Bldg.)
LEGEND:
E - EBASCO CE - COMBUSTION ENGINEERING
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 2 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS
DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (A) Instrument Location Arrangement Drawings (Cont'd)
LOU-5817) G-391) Dwg's for HVAC System 70% @ 9/77 1 3 11-81 E Turbine Building, HWAC System, Instrumentation 2 2 12-80 E No Turbine Building, HVAC System Instrumentation 3 3 08-81 E No Reactor & Fuel Handling Bldg, HVAC Instrumentation 4 4 01-82 E No Reactor Auxiliary Building, HVAC Instrumentation 5 4 11-81 E No Reactor Auxiliary Building, HVAC Instrumentation 6 3 11-80 E No Reactor Auxiliary Building, HVAC Instrumentation 7 3 11-81 E No HVAC Instrumentation Location Arrangement
(B) Instrument Schematics and Logic Diagrams - Safety Related
B-431 Cover 0 3-8-73 E No Instrument Schematics and Logic Diagrams A 1 1-18-80 E No Organization of Drawing B 6 4-7-80 E No Index c 6 4-7-80 E No Index D 4 4-7-80 E No Index E 6 4-7-80 E No Index F 1 4-7-80 E No Index 0.1 0 9-30-71 E No Pipe, Valves & Instrument Symbols (Cover Sheet) 0.2 4 12-19-79 E No Instrumentation - Symbols 0.3 3 7-23-79 E No Instrumentation - Functions & Meanings or Identification Letters 0.4 3 12-19-79 E No Line Designations 0.5 3 1-18-80 E No Line System Description and Symbols (Abbreviations) 0.6 4 1-18-80 E No Abbreviations (Continued) 0.7 4 1-23-80 E No Valves Symbols 0.8 3 12-19-79 E No Valves Symbols 0.9 1 12-19-79 E No Pipe or Equipment Accessories Symbols 0.10 2 12-19-79 E No Equipment Symbols 0.11 1 12-19-79 E No Graphical Symbols for Logic Diagrams 3AS 4 12-19-79 E No Contaiment Spray 4AS 4 12-19-79 E No Containment Spray 5AS 3 12-19-79 E No Contaiment Sump Instruments & Leak Detection 65 3 12-19-79 E No Containment Pressure 145 2 12-19-79 E No Main Steam Isolation Valve Steam Generator - 1 155 2 12-19-79 E No Main Steam Isolation Valve Steam Generator - 2 35ASI 1 12-19-79 E No Emergency Steam Generator Feedwater System 35AS2 2 6-11-80 E No Emergency Steam Generator Feedwater System 95AS 3 12-19-79 E No Condensate Storage Pool Instrumentation
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 3 of 47) Revision 13-B (01/05) ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS
DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (B) Instrument Schematics and Logic Diagrams - Safety Related
114AS 2 6-19-78 E No Shutdown Heat Exchangers 115AS 2 6-19-78 E No Component Cooling Water Supply to & Return from containment 118AIS 3 12-19-79 E No Containment Fan Cooler Unit System Instrumentation 118A2S 3 12-19-79 E No Containment Fan Cooler Unit System Instrumentation 120AS 2 6-11-80 E No Component Cooling Water Pump Schematic 121AS 3 11-6-80 E No Component Cooling Water Surge Tank & Return Headers 122AS 3 6-11-80 E No Component Cooling Water System A (Redrawn) 123AS 3 6-11-80 E No Component Cooling Water System B (Redrawn)
B-431 124AS 3 6-11-80 E No Component Cooling Water Header Instrumentation 125AS 3 6-11-80 E No Component Cooling Water to Control Room Chillers 126AS 1 6-19-78 E No Component Cooling Water to Diesel Gen Heat Exchangers 153S 1 12-19-79 E No Diesel Oil System Inst. (Diesel Oil Feed Tank "A") 154S 1 12-19-79 E No Diesel Oil System Inst. (Diesel Oil Feed Tank "B") 220S 0 8-29-78 E No Miscellaneous Instrumentation 221S 1 8-29-78 E No RAB-CCW Pump Room "A", Air Handling Unit AH-10 (3A-SA) Instrumentation 222S 1 8-29-78 E No RAB-CCW Pump Room "B", Air Handling Unit AH-10 (3B-SB) Instrumentation 223S 1 8-29-78 E No RAB-Emerg. FW Pump "A" Room "A", Air Handling Unit AH-17 (3A-SA) Instrumentation 224S 1 8-29-78 E No RAB-Emerg. FW Pump "B" Room "B", Air Handling Unit AH-17 (3B-SB) Instrumentation 225S 1 8-29-78 E No RAB-Charging Pump "A" Room "A", Air Handling Unit AH-18 (3A-SA) Instrumentation 226S 1 8-29-78 E No RAB-Charging Pump "B" Room "B", Air Handling Unit AH-18 (3B-SB) Instrumentation 229S 2 2-12-79 E No RAB-Shutdown Ht. Exch. "B" Room "B", Air Handling Unit AH-3 (3B-SB) Instrumentation 230S 1 8-29-78 E No RAB-Sbuidown Ht. Exch "A" Room "A", Air Handling Unit AH-3 (3A-SA) Instrumentation 231S 1 8-29-78 E No RAB-CCW Pump Ht. Exch. "B" Room "B", Air Handling Unit AH-24 (3B-SB) Instrumentation 232S 1 8-29-78 E No RAB-CCW Pump Ht. Exch. "A" Room "A", Air Handling Unit AH-24 (3A-SA) Instrumentation 233S 2 2-12-79 E No RAB-Safeguard Pumps Room "A", Air Handling Unit AH-2 (3A-SA) Instrumentation 234S 2 2-12-79 E No RAB-Safeguard Pumps Room "A", Air Handling Unit AH-2 (3C-SA) Instrumentation
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 4 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS
DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (B) Instrument Schematics and Logic Diagrams - Safety Related
B-431 235S 2 2-12-79 E No RAB-Safeguard Pumps Room "B", Air Handling Unit AH-2 (3B-SB) Instrumentation 236S 2 2-12-79 E No RAB-Safeguard Pumps Room "B", Air Handling Unit AH-2 (3D-SB) Instrumentation 237S 2 2-12-79 E No RAB-Safeguard Pump Room "A", Air Handling Unit AH-21 (3-SAB) instrumentation 238S 1 8-29-78 E No RAB-Control Room Equipment Room, Air Handling Unit AH-26 (3A-SA) Instrumentation 239S 1 8-29-78 E No RAB-Control Room Equipment Room, Air Handling Unit AH-26 (3B-SH) Instrumentation 240S 1 8-29-78 E No RAB-CCW Pump Room "A/B", Air Handling Unit AH-20 (3B-SAB) Instrumentation 241S 1 8-29-78 E No RAB-CCW Pump Room "A/B", Air Handling Unit AH-20 (3A-SAB) Instrumentation 242S 1 8-29-78 E No RAB-Charging Pump Room "A/B", Air Handling Unit AH-22 (3A-SAB) Instrumentation 243S 1 8-29-78 E No RAB-Charging Pump Room "A/B", Air Handling Unit AH-22 (3B-SAB) Instrumentation 245S 2 12-14-79 E No Control Atmos Release System (CARS) Instrumentation 246S 1 8-29-78 E No Switchgear Area Ventilation System AH-25 (3A-SA) 247S 1 8-29-78 E No Switchgear Area Ventilation System AH-25 (3B-SB) 248S 1 8-29-78 E No Battery Rm Exhaust Fans 251S 2 12-14-79 E No Emerg Diesel Gen Rm "A" Vent Sys Fan E-28 (3A-SA) 252S 2 12-19-79 E No Emerg Diesel Gen Rm "B" Vent Sys Fan E-28 (38-SB) 253S 2 12-14-79 E No Water Chiller WC-1 (3A-SA) 254S 2 12-14-79 E No Water Chiller WC-1 (3B-SB) 255S 2 12-14-79 E No Water Chiller WC-1 (3C-SA/B) 256S 2 12-14-79 E No Chiller Water Sys Isol Valves 260S 1 12-06-78 Control Room, Conference Room & Kitchen Exhaust Fan 261S 1 12-06-78 E No Control Rm Toilet Exh Fans E-34 (3A-SA) & E-34 (3B-SB) 262S 1 12-14-79 E No Swgr Area Vent System AH-30 (3A-SA) 263S 2 12-14-79 E No Swgr Area Vent System AH-30 (3B-SB) 264S 3 9-18-81 E No Cont Rm Emerg Filtration Unit S-8 (3A-SA) 265S 2 9-18-81 E No Cont Rm Emerg Filtration Unit S-8 (3B-SB) 266S 1 12-6-78 E No Cont Rm Air Handling Unit AH-12 (3A-SA) G(DRN 01 758, R11 A) 267S 2 12-14-79 E No Cont Rm Air Handling Unit AH-12 (3B-SB) 0(DRN 01 758, R11 A) 268S 3 9-18-81 E No SH1eld Bldg Vent System E-17 (3A-SA)
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 5 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS
DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (B) Instrument Schematics and Logic Diagrams - Safety Related
B-431 269S 3 9-18-81 E No SH1eld Bldg Vent System E-17 (3B-SB) 270S 1 2-12-79 E No RAB HVAC Equip Rm Vent System AH-13 (3A-SA) & E-41 (3A-SA) 271S 1 2-12-79 E No AH-13 (3A-SA) Rm Vent System AH-13 (3B-SB) & E-41 (3B-SB) 272S 0 8-29-78 Fuel Handling Building Normal Supply Fan AH-14 (3) 274S 2 9-18-81 E No FHB Emerg Filtration Unit E-35 (3A-SA) 275S 2 9-18-81 E No FHB Emerg Filtration Unit E-35 (3B-SB) 276S 0 8-29-78 Fuel Handling Building Mechanical Equipment Room Exhaust Fans E-21 (3A-SA) & E-21 (3B-SB) 283S 1 9-8-81 E No Containment Vacuum Relief 284S 0 8-29-78 E No Containment Fan Clr AH-1 (3C-SA) & (3A-SA) 285S 0 8-29-78 E No Containment Fan Clr AH-1 (3D-SB) & (3B-SB) 289S 0 8-29-78 E No Battery Rooms Fan Room Fans E-52 (3A-SA) & (3B-SB) 295S 2 9-8-81 E No Containment Purge System 296S 1 9-8-81 E No CVAS Exhaust Fan E-23 (3A-SA) 297S 2 9-8-81 E No CVAS Exhaust Fan E-23 (3B-SB) 298S 1 12-19-79 1 No CVAS Isolation Valves 299S 1 12-19-79 E No CVAS Isolation Valves
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 6 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS
DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (B) Instrument Schematics and Logic Diagrams - Safety Related
B-431 1 2 12-14-79 E No Containment Spray 2A 3 12-19-79 E No Refueling Water Storage Pool Makeup Control 11 1 12-14-79 E No Main Steam System 12 4 12-19-79 E No Main Steam From Steam Generator - 1 13A 4 12-19-79 E No Main Steam From Steam Generator - 2 16A 4 12-19-79 E No Steam to Steam Generator Feedwater Pump Turbines 17A 3 12-19-79 E No Main Steam to H.P. Turbine 18A 2 6-19-78 E No Steam Dump Lines 19 4 2-12-79 E No Main Steam to Moisture Separator Reheaters 24 0 12-14-79 E No Auxiliary Steam System 25 1 12-19-79 E No Auxiliary Steam System 26 1 12-19-79 E No Auxiliary Steam System 31 1 12-14-79 E No Feedwater System 32A-1 3 3-19-80 E No Feedwater Pump "A" Flow System 32A-2 1 12-19-79 E No Feedwater Pump "A" Flow System 33A-1 1 3-19-80 E No Feedwater Pump "B" Flow System 33A-2 1 12-19-79 E No Feedwater Pump "B" Flow System 34 4 12-19-79 E No Feedwater to H.P. Heater IA/LB/IC 46 1 12-14-79 E No Extraction Steam System 47A 4 12-19-79 E No Extraction Steam From H.P. Turbine to H.P. Heaters IA/IB/IC 48A 4 12-19-79 E No Cold Rebeat Steam from H.P. Turbine to L.P. Heaters 2A/2B/2C 49 2 6-19-78 E No Steam to L.P. Turbines A/B/C 50 3 6-19-78 E No Condenser Vacuum 51A 4 12-19-79 E No Extraction Steam From L.P. Turbine to I.P. Heater 3A 52A 4 12-19-79 E No Extraction Steam From L.P. Turbine to I.P. Heater 3B 53A 4 12-19-79 E No Extraction Steam From L.P. Turbine to I.P. Heater 3C 54A 4 12-19-79 E No Extraction Steam From L.P. Turbine to I.P. Heater 4A 55A 4 12-19-79 E No Extraction Steam From L.P. Turbine to I.P. Heater 4B 56A 4 12-19-79 E No Extraction Steam from L.P. Turbine to I.P. Heater 4C 58 0 6-19-78 E No L.P. Heaters 5 & 6 Shell Pressure 59 0 6-19-78 E No Cold Reheat from H.P. Turbine to Moisture Separator Reheater 66 1 12-14-79 E No Heater Drain System 67A 4 12-19-79 E No Moisture Separator "A" Drain System Instrumentation 68A 4 12-19-79 E No Moisture Separator "B" Drain System Instrumentation 69 3 12-19-79 E No Heater IA/IB/IC Drain System Instrumentation 70A 5 8-10-81 E No Heater 2A & Drain Pump System Instrumentation 71A 5 8-10-81 E No Heater 2B & Drain Pump System Instrumentation 72A 5 8-10-81 E No Heater 2C & Drain Pump System Instrumentation
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 7 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS
DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (B) Instrument Schematics and Logic Diagrams - Non-Safety Related
B-431 73 3 12-19-79 E No Heater 3A/3B/3C Drain System Instrumentation 74 3 12-19-79 E No Heater 4A/4B/4C Drain System Instrumentation 75 3 12-19-79 E No Heater SA/5B/SC Drain System Instrumentation 76 3 12-19-79 E No Heater 6A/6B/6C Drain System Instrumentation 77 2 2-12-80 E No M.S.R. "A" Shell Drain List 78 2 2-12-80 E No M.S.R. "B" Shell Drain List 86 1 12-14-79 E No Condensate System 87 4 12-19-79 E No Condensate To and From Gland Steam Condenser 88A 4 12-19-79 E No Condenser Hotwell Level Control 89 4 2-12-80 E No Condensate to and from Heaters in Condenser Neck 91A 3 12-19-79 E No Gland Seal Leakoff Tank Level Control 92 3 2-12-80 E No Condensate to and from IP Heaters 93A 4 12-19-79 E No Condensate Pumps 94A 3 12-19-79 E No Primary Water & CNDS Storage Tank System 101 1 12-14-79 E No Turbine Closed Cooling Water System 102A 3 2-12-80 E No Surge Tank & Turbine Cooling Water Pumps Instrumentation System 103 3 2-12-80 E No Turbine Cooling Water Heat Exchangers Instrumentation System 104 2 12-19-79 E No Cooling Water to Turbine L.O. Coolers 105 0 6-19-78 E No Cold Gas System Generator Cooling 106 1 12-19-79 E No Steam Generator Feed Pump Turbine Coolers "A & B" 107 0 6-19-78 E No Generator Stator Water Coolers 108 0 6-19-78 E No Turbine Electro Hydraulic Fluid Cooling 111 1 12-14-79 E No Component Closed Cooling Water System 112 3 12-19-79 E No Cooling Water System for Pump Coolers (TYP) 113 1 6-19-78 E No Waste and Boric Acid Concentrator 116A 3 12-19-79 E No Reactor Coolant Pumps Coolers 117 4 12-19-79 E No Letdown Heat Exchanger Cooling Water System 119 4 12-19-79 E No Fuel Pool Heat Exchanger Cooling Water System 128 1 12-14-79 E No Firewater and Domestic Water System 129A 2 7-23-79 E No Firewater Storage System 136 1 12-14-79 E No Lube Oil System 146 1 12-14-79 E No Instrument and Station Air System 147 2 7-23-79 E No Station Air System 148 2 7-23-79 E No Instrument Air System 152 1 12-14-79 E No Emergency Diesel Generator Oil System 156 1 12-14-79 E No Condenser Cooling Water System 157A 2 12-19-79 E No Circulating Water Pumps Schematic 158 1 6-19-78 E No Cooling Water to Turbine Closed Cooling Water Heat Exchangers 159A 1 6-19-78 E No CW Air Evacuation Pumps Schematic
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 8 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS
DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (B) Instrument Schematics and Logic Diagrams - Non-Safety Related
B-431 160 1 6-19-78 E No Cooling Water to and From Condenser "A" 161 1 6-19-78 E No Cooling Water to and From Condenser "B" 162 1 6-19-78 E No Cooling Water to and From Condenser "C" 163A-1 0 6-30-78 E No Screen Wash System 163A-2 0 6-30-78 E No Screen Wash System 171 1 12-14-79 E No Gas System 181 1 12-14-79 E No Blowdown System 182 0 9-19-78 E No Steam Generator Blowdown System Inst. 183 1 12-19-79 E No Blowdown Tank & Blowdown Pumps "A & B" 184 1 12-19-79 E No Blowdown Heat Exchangers "A & B" 185 1 12-19-79 E No Blowdown Demineralizer System 190 0 6-19-78 E No Sump Pumps Level Instruments 191 1 12-14-79 E No Sampling Systems 201 1 12-14-79 E No Containment Air 202 2 12-19-79 E No Containment Radiation Monitoring 211 1 12-14-79 E No HVAC System 212 1 12-14-79 E No Supplementary Chiller Cooling Tower and Chiller Water Control 213 1 12-14-79 E No Supplementary Chiller Cooling Tower and Chiller Water Control 214 1 12-14-79 E No Supplementary Chilled Water Control 215 1 12-14-79 E No Supplementary Chilled Water Cooling Tower Makeup Control 217 1 5-05-81 E No Condensate Pump "A" Motor Exhaust Fans E-53 (3A&3B) 218 0 8-29-78 E No Condensate Pump "B" Motor Exhaust Fans E-54 (3&3B)
219 0 8-29-78 E No Condensate Pump "C" Motor Exhause Fans E-55 (3A&3B) 227 1 8-29-78 E No RAB Blowdown Filter "B" Room "B" Air Handling Unit AH-28 (3) Instrumentation 228 8-29-78 E No RAB Blowdown Filter "A" Room "A", Air Handling Unit AH-27 (3) Instrumentation 244 2 12-14-79 E No Reactor Cavity Cooling System Instrumentation S-2(3A) & S-2 (3B) 249 2 7-23-79 E No Turbine Area Supply Fan S-10 (3) 250 2 7-23-79 E No Typical Turbine Area Exhaust Fan 257 2 12-14-79 E No Annulus Negative Pressure System 258 0 9-29-78 E No Airborne Radioactivity Removal System Fan E-13 (3A) 259 0 9-29-78 E No Airborne Radioactivity Removal System Fan E-13 (3B) 273 0 9-29-78 E No Fuel Handling Building Normal Exhaust Fans E-20 (3A) & E-20(3B) 277 0 9-29-78 E No CEDM Fans (Inlet) 278 0 2-12-78 E No CEDM Fans E-16(3A), E-16(3B), E-16(3C) & E-16(3D) 279 0 9-29-79 E No Reactor Aux. Bldg. AH-5(3) & E-47(3)
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 9 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS
DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (B) Instrument Schematics and Logic Diagrams - Non-Safety Related
B-431 280 0 9-29-78 E No Reactor Aux Bldg. AH-8(3) 281 0 9-29-78 E No RAB Smoke Purge Fans E-48(3) & E-50(3) 282 0 9-29-78 E No Cable Vault Area Exhaust Fan E-49(3) 286 0 9-29-78 E No Containment Fan Coolers Condensate Flow Detection 287 0 2-25-81 E No Turbine Bldg. Swgr Unit AH-15(3) 288 0 2-25-81 E No Turbine Bldg. Swgr Unit AH-29(3) 290 0 9-29-78 E No Computer Room Supplemental Air Handling System AH-31(3) 291 0 10-16-78 E No Hot Machine Shop Ventilation System 292 2 12-19-78 E No Decontamination Facility Ventilation System 293 1 10-16-78 E No RAB Normal Supply Fans 294 1 12-19-79 E No RAB Normal Exhaust Fans
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 10 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS
DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams - Safety Related
B-424 I Open 9-82 E No Index II Open 9-82 E No Index III Open 9-82 E No Index IV Open 9-82 E No Index V Open 9-82 E No Index VI Open 9-82 E No Index VII Open 9-82 E No Index VIII Open 9-82 E No Index IX Open 9-82 E No Index X Open 9-82 E No Index XI Open 9-92 E No Index XII Open 9-82 E No Index XIII Open 9-82 E No Index XIV Open 9-82 E No Index XV Open 9-82 E No Index XVI Open 9-82 E No Index XVII Open 9-82 E No Index XVIII Open 9-82 E No Index XIX Open 9-82 E No Index XX Open 9-82 E No Index XXI Open 9-82 E No Index XXII Open 9-82 E No Index XXIV 1 9-82 E No List of Abbreviations XXV 1 9-82 E No List of Abbreviations XXVI 1 9-82 E No List of Abbreviations XXVII 1 9-82 E No List of Abbreviations XXVIII 1 9-82 E No List of Abbreviations XXIX 1 9-82 E No List of Abbreviations XXX 1 9-82 E No List of Abbreviations XXXI 1 9-82 E No List of Abbreviations XXXII 1 9-82 E No List of Abbreviations XXXIII 0 9-82 E No Device Numbers and Definitions XXXIV 0 9-82 E No Device Numbers and Definitions XXXV 0 9-82 E No Device Numbers and Definitions XXXVI 0 9-82 E No Device Numbers and Definitions XXXVII 0 9-82 E No Device Numbers and Definitions XXXVIII 0 9-82 E No Device Numbers and Definitions XL 1 10-21-78 E No Graphical Symbols XLI 0 10-21-78 E No Graphical Symbols
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 11 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS
DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams - Safety Related (Cont'd)
B-424 XLII 0 2-28-80 E No Typical Breaker Closing Ciruits XLIII 2 4-11-80 E No Typical Plug Connection Diagram XLIV 0 2-28-80 E No Isolation Relays - Typcial Wiring Diagrams IS 9 9-82 E No CEA Drive MG Set A Supply BKR 3S 10 8-82 E No CFA Drive MG Set B Supply BKR 5S 7 10-27-80 E No Reactor Trip Bkr TCBI 6S 6 10-27-80 E No Reactor Trip Bkr TCH2 7S 7 9-82 E No Reactor Trip Bkr TCB3 8s 6 10-27-80 E No Reactor Trip Bkr TCB4 9s 5 10-27-80 E No Reactor Trip Bkr TCB5 10s 5 10-27-80 E No Reactor Trip Bkr TCB6 11s 5 10-27-80 E No Reactor Trip Bkr TCB7 12S 5 10-27-80 E No Reactor Trip Bkr TCB8 13S 7 10-27-80 E No Reactor Trip Bkr TCB9 21S 10 9-82 E No CEA No. 1 22S 9 9-82 E No CEA No. 2 23S 9 9-82 E No CFA No. 3 24S 8 9-82 E No CEA No. 4 25S 9 9-82 E No CEA No. 5 26S 9 9-82 E No CEA No. 6 27S 8 9-82 E No CEA No. 7 28S 8 9-82 E No CEA No. 8 29S 9 9-82 E No CEA No. 9 30S 8 9-82 E No CEA No. 10 31S 8 9-82 E No CFA No. 11 32S 8 9-82 E No CEA No. 12 33S 8 9-82 E No CFA No. 13 34S 8 9-82 E No CEA No. 14 35S 9 9-82 E No CEA No. 15 36S 8 9-82 E No CEA No. 16 37S 9 9-82 E No CEA No. 17 38S 8 9-82 E No CEA No. 18 39S 8 9-82 E No CEA No. 19 40S 8 9-82 E No CEA No. 20 41S 8 9-82 E No CEA No. 21 42S 8 9-82 E No CEA No. 22 43S 9 9-82 E No CEA No. 23 44S 8 9-92 E No CEA No. 24 45S 8 9-82 E No CEA No. 25
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 12 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS
DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams - Safety Related (Cont'd)
B-424 46S 9 9-82 E No CEA No. 26 47S 8 9-82 E No CEA No. 27 48S 8 9-82 E No CEA No. 28 49S 8 9-82 E No CEA No. 29 50S 8 9-82 E No CEA No. 30 51S 8 9-82 E No CEA No. 31 52S 8 9-82 E No CEA No. 32 53S 9 9-82 E No CEA No. 33 54S 8 9-82 E No CEA No. 34 55S 8 9-82 E No CEA No. 35 56S 8 9-82 E No CEA No. 36 57S 9 9-82 E No CEA No. 37 5as 9 9-82 E No CEA No. 38 59S 9 9-82 E No CEA No. 39 60S 8 9-82 E No CEA No. 40 61S 8 9-82 E No CFA No. 41 62S 8 9-82 E No CEA No. 42 63S 8 9-82 E No CEA No. 43 64S 9 9-82 E No CEA No. 44 65S 8 9-82 E No CEA No. 45 66S 8 9-82 E No CEA No. 46 67S 8 9-82 E No CEA No. 47 68S 8 9-82 E No CEA No. 48 69S 8 9-82 E No CEA No. 49 70S 9 9-82 E No CEA No. 50 71S 8 9-82 E No CRA No. 51 72S 10 9-82 E No CEA No. 52 73S 10 9-82 E No CEA No. 53 74S 9 9-82 E No CEA No. 54 75S 9 9-82 E No CEA No. 55 76S 9 9-82 E No CEA No. 56 77S 8 9-82 E No CEA No. 57 78S 8 9-82 E No CEA No. 58 79S 8 9-82 E No CEA No. 59 80S 8 9-82 E No CEA No. 60 81S 8 9-82 E No CEA No. 61 82S 8 9-82 E No CEA No. 62 83S 8 9-82 84S 8 9-82 E No CEA No. 64
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) (DRN 02-85, R11-A; 04-1444, R13-B) Start of historical information. (DRN 04-1444, R13-B)
WSES-FSAR-UNIT 3 (DRN 02-85, R11-A) TABLE 1.7-1 (Sheet 13 of 47) Revision 307 (07/13)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS
DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams - Safety Related (Cont'd)
B-424 85S 8 9-82 E No CEA No. 65 86S 9 9-82 E No CEA No. 66 87S 8 9-82 E No CEA No. 67 88S 8 9-82 E No CEA No. 68 89S 8 9-82 E No CEA No. 69 90S 8 9-82 E No CEA No. 70 91s 8 9-82 E No CEA No. 71 92S 8 9-82 E No CEA No. 72 93S 8 9-82 E No CEA No. 73 94S 8 9-82 E No CEA No. 74 95S 9 9-82 E No CEA No. 75 96S 8 9-82 E No CEA No. 76 97S 9 9-82 E No CEA No. 77 98S 8 9-82 E No CEA No. 78 99S 8 9-82 E No CEA No. 79 100S 8 9-82 E No CEA No. 80 101S 8 9-82 E No CEA No. 81 102S 8 9-82 E No CEA No. 82 103S 8 9-82 E No CEA No. 83 104S 9 9-82 E No CEA No. 84 105S 8 9-82 E No CEA No. 85 106S 9 9-82 E No CEA No. 86 107S 8 9-82 E No CEA No. 87 (EC-2800, R307)
(EC-2800, R307) 113S 4 9-82 E No Excore Neutron Flux Protective Channel A 114S 4 9-82 E No Excore Neutron Flux Protective Channel B 115S 4 9-82 E No Excore Neutron Flux Protective Channel C 116S 4 9-82 E No Excore Neutron Flux Protective Channel D 119S 9 9-82 E No Incore Neutron Flux Det 93A-L16 & 93A-G18 Incore Neutron Flux 120S 9 9-82 E No Det 93A-E20 & 93A-GI3 Incore Neutron Flux 121S 8 9-82 E No Det 95A-A14 & 95A-C9 Incore Neutron Flux
(DRN 04-1444, R13-B) End of historical information. (DRN 04-1444, R13-B)
G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 14 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS
DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams - Safety Related (Cont'd)
B-424 122S 8 9-82 E No Det 95A-E6 & 96A-C2 Incore Neutron Flux 123S 8 9-82 E No Det 96A-C4 & 98A-T2 Incore Neutron Flux 124S 8 9-82 E No Det 98A-19 & 98AR4 Incore Neutron Flux 125S 8 9-82 E No Det 100A-T16 & 101A-WI8 Incore Neutron Flux 126S 8 9-82 E No Det 92B-R18 & 92B-120 Incore Neutron Flux 127S 8 9-82 E No Det 94B-E16 & 94B-C18 Incore Neutron Flux 128S 8 9-82 E No Det 96B-E2 & 97B-G6 129S 8 9-82 E No Incore Neutron Flux Det 97B-R2 & 97B-G4 130S 8 9-82 E No Incore Neutron Flux Det 99B-W4 & 99B-T9 131S 8 9-92 E No Incore Neutron Flux Det 99B-R13 & 99B-T6 132S 8 9-92 E No Incore Neutron Flux Det 100B-Y8 & 101B-T20 133S 9 9-82 E No Incore Neutron Flux Det 92C-C20 & 93C-G16 134S 8 9-82 E No Incore Neutron Flux Det 93C-L13 & 94C-C9 135S 8 9-82 E No Incore Neutron Flux Det 94C-C16 & 9SC-E9 136S 8 9-82 E No Incore Neutron Flux Det 96C-E4 & 96C-Ag 137S 8 9-82 E No Incore Neutron Flux Det 97C-L2 & 98C-R6 138S 8 9-82 E No Incore Neutron Flux Det 99C-W6 & 100C-W13 139S 8 9-82 E No Incore Neutron Flux Det 100C-T13 & I0IC-T18 140S 8 9-82 E No Incore Neutron Flux Det 92D-L1B & 92D-R16
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 15 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS
DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams - Safety Related (Cont'd)
B-424 141S 8 9-82 R No Incore Neutron Flux Det 93D-E18 & 94D-EI3 142S 8 9-82 E No Incore Neutron Flux Det 95D-C13 & 96D-C6 143S 8 9-82 E No Incore Neutron Flux Det 97D-L4 & 98D-T4 144S 8 9-82 E No Incore Neutron Flux Det 98D-L6 & 99D-R9 145S 8 9-82 E No Incore Neutron Flux Det 1OOD-W9 & 101D-Y14 146S 8 9-82 E No Incore Neutron Flux Det 101D-W16 & 101D-R20 147S 5 9-82 E No Incore Thermocouples SH. 1 148S 5 9-9-80 E No Incore Thermocouples SH. 2 149S 5 9-82 E No Incore Thermocouples SH. 3 150S 5 9-9-80 E No Incore Thermocouples SH. 4 151S 5 9-9-80 E No Incore Thermocouples SH. 5 152S 5 9-9-80 E No Incore Thermocouples SH. 6 153S 5 9-9-80 R No Incore Thermocouples SH. 7 154S 5 9-9-80 E No Incore Thermocouples SH. 8 160S 8 9-82 E No ESFAS Auxiliary Relay Cab A Alarms Sh. 1 162S 2 10-4-78 E No ESFAS Auxiliary Relay Cab A Test Module 165S 3 9-82 E No ESFAS Auxiliary Relay Cab B Alarms Sh. 1 167S 2 3-9-79 E No ESFAS Auxiliary Relay Cab B Test Module 170S 8 9-82 E No ESFAS Auxiliary Relays 176S 8 9-26-79 E No Auxiliary Protective Cab Computer & Alarm Channel A 177S 10 9-82 E No Auxiliary Protective Cab Computer & Alarm Channel B 178S 10 9-82 E No Auxiliary Protective Cab Computer & Alarm Channel C 179S 9 9-82 E No Auxiliary Protective Cab Computer & Alarm Channel D 180S 2 9-82 E No Auxiliary Protective Cab Interconnections 188S 4 9-82 E No Control Board Mounted Nuclear Instrumentation Sh. 2 189S 7 9-82 E No Control Board Mounted Nuclear Instrumentation Sh. 3
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 16 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS
DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams - Safety Related (Cont'd)
B-424 19is 3 3-22-79 E No Control Board Mounted Nuclear Instrumentation Sh. 5 193S 2 3-27-79 E No Control Board Miscellaneous Instrumentation Sh. 2 194S 4 4-3-79 E No Control Board Miscellaneous Instrumentation Sh. 3 195S 4 9-82 E No Control Board Miscellaneous Instrumentation Sh. 4 199S 7 9-82 E No Power Supplies Pnl's CP-25, 26, 27, 28, 29, 30, 31 201S 9 9-82 E No RCS Loop 1 Temp (Hot Leg) 202S 9 9-82 E No RCS Loop 2 Temp (Hot Leg) 207S 9 9-82 E No RCS Loop 1 Temp (Cold Leg) 208S 9 9-82 E No RCS Loop 2 Temp (Cold Leg) 210S 7 9-82 E No Steam Gen No. 1 Press Sh. 1 211S 4 9-82 E No Steam Gen No. 1 Press Sh. 2 212S 7 9-82 E No Steam Gen No. 1 Level Sh. 1 213S 4 9-82 E No Steam Gen No. 1 Level Sh. 2 215S 7 9-82 E No Steam Gen No. 2 Press Sh. 2 216S 4 9-82 E No Steam Gen No. 2 Press Sh. 2 217S 4 9-82 E No Steam Gen No. 2 Level Sh. 1 218S 5 9-82 E No Steam Gen No. 2 Level Sh. 2 225S 11 9-82 E No Reactor Coolant Pump 1A Alarms and Computer Input Sh 2 226S 4 9-82 E No Reactor Coolant Pump 1A Protective Speed 235S 14 9-82 E No Reactor Coolant Pump 1B Alarms and Computer Inputs Sh 2 236S 3 9-82 E No Reactor Coolant Pump 1B Protective Speed 245S 12 9-82 E No Reactor Coolant Pump 2A Alarms and Computer Inputs Sh 2 246S 3 9-82 E No Reactor Coolant Pump 2A Protective Speed 255S 14 9-82 E No Reactor Coolant Pump 2B Alarms and Computer Inputs Sh 2
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 17 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS
DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE G (DRN 01 758, R11 A) (C) Control Wiring Diagrams - Safety Related (Cont'd) B-424 256S 3 9-82 E No Reactor Coolant Pump 2B Protective Speed 262S 6 9-82 E No Loose Parts Monitoring System Sh 3 266S 6 9-82 E No Pressurizer Press (Narrow Range) Sh 3 267S 6 9-82 E No Pressurizer Press (Wide Range) Sh 4 268S 5 9-82 E No Pressurizer Press (Wide Range) Sh 5 269S 6 9-82 E No Pressurizer Press Sh 6 270S 10 9-82 E No Pressurizer Press (Wide Range) Sh 7 273S 9 9-82 E No Pressurizer Level Sh 1 275S 6 9-82 E No Pressurizer Level Sh 3 278S 2 9-82 E No Subcooled Marg. Monitor Channel A 279S 2 9-82 E No Subcooled Marg. Monitor Channel B 294S 3 9-82 E No Pressurizer Aux Spray Valve ICH-E2505A 295S 4 9-82 E No Pressurizer Aux Spray Valve ICH-E2505B 300S 8 9-82 E No Letdown Stop V& ICH-FIS16A/B 301S 7 9-82 E No Letdown Cont Isol Valve ICH-F25OIA/B 302S 6 8-27-80 E No Letdown Cont Isol Valve 2CH-FI51BA/B 325S 6 9-82 E No RCP Bleedoff Cont Isol Valves 2CH-FIS12A/B & 2CH-FIS13A/B 327S 5 9-82 E No Volume Control Tk Disch Valve 2CH-VI23A/B 337S 4 2-7-79 E No Boric Acid Gravity Feed Valve 3CH-VI06A 338S 3 11-28-79 E No CVCS System A Heat Tracing 340S 7 12-18-79 E No Boric Acid Makeup Tk B Heater 2-1 341S 8 1-17-80 E No Boric Acid Makeup Tk B Heater 2-2 342S 4 2-26-79 E No Boric Acid Gravity Feed Valve 3CH-VlO7B 343S 4 9-82 E No CVCS System B Heat Tracing 345S 6 9-92 E No Boric Acid Pump A 346S 5 9-82 E No Boric Acid Pump A Recirc Valve 3CH-Fl7OA 350S 6 9-82 E No Boric Acid Pump B 351S 4 5-16-79 E No Boric Acid Pump B Recirc Valve & 3CH-FI71B 357S 5 9-82 E No Reactor Makeup Stop Valve 3CH-FI17A/B 358S 6 9-92 E No Reactor Makeup Bypass Valve 3CH-VI12A/B
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 18 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams - Safety Related (Cont'd)
B-424 365S 5 9-82 E No Charging Pump A 366S 8 9-82 E No Charging Pump A Seal Lube Pump 370S 4 9-92 E No Charging Pump A/B 371S 8 9-82 E No Charging Pump A/B Seal Lube Pump 375S 4 9-82 E No Charging Pump B 376S 8 9-82 E No Charging pump B Seal Lube Pump 381S 8 9-82 E No Charging Line to loop 1A; Shutoff Valve ICH-E2503A 382S 8 9-82 E No Charging Line to loop 2A; Shutoff Valve ICH-E2504B 402S 4 9-82 E No Reactor Drain Tk Cont Isol Valves 2BM-FIOBA/B & 2BM-FI09A/B 490S 8 9-82 E No Refueling Water Tk Level Sh 1 492S 5 11-19-80 E No Refueling Water Storage Pool Outlet Valve 2SI-L103A 493S 7 9-82 E No SIS Sump Isol Valve 2SI-L101A 495S 6 9-82 E No Refueling Water Storage Pool Outlet Valve 2SI-LI04D 496S 6 9-82 E No Cont Sump Isol Valve 2SI-L102B 498S 6 9-82 E No Cont Isol from SI TK Drain To RWP Valve 2SI-FI561A/B 500S 8 9-82 E No High Press Safety inj Pump A 502S 4 9-82 E No High Press SI Header A Orifice Bypass Valve 2SI-VIS34 503S 5 9-82 E No Reactor Coolant Loop I Hot Leg Flow Contr Valve 2SI-VI556 504S 5 9-82 E No Reactor Coolant Loop I Hot Let Inj Isol Valve 251-VIS57 507S 5 9-82 E No High Press Safety Inj Pump A/B 509S 7 9-82 E No High Pressure Safety Inj Pump B 511S 3 9-82 E No High Pressure SI Header B Orifice Bypass Valve 2SI-V811B 512S 5 9-82 E No Reactor Coolant Loop 2 Hot Leg Flow Contr Valve 2SI-Vl559 513S 4 9-82 E No Reactor Coolant Loop 2 Hot Let Inj Isol Valve & 2SI-VI558 516S 5 9-82 E No Safety Inj Pumps A Min Flow Isol Valve 2SI-V809A
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 19 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams - Safety Related (Cont'd)
B-424 517S 5 9-82 E No Safety Inj Pumps A Min Flow Isol Valve 2SI-VB10A 518S 5 9-82 E No Safety Inj Pumps B Min Flow Isol Valve 2SI-V801B 519S 6 9-82 E No Safety Inj Pumps B Min Flow Isol Valve 2SI-V802B 520S 1 9-82 E No LPSI Pumps Min. Flow isolation Valves 2ST-EI587A and 2SI-EI588B 521S 5 9-82 E No High Press SI Flow Cont Valve 2ST-VI550AI 522S 5 9-82 E No High Press Sl Flow Cont Valve 2SI-VIS46A2 523S 5 9-82 E No High Press Sl Flow Cont Valve 2SI-VI542A3 524S 5 9-82 E No High Press Sl Flow Cont Valve 2SI-VI548A4 525S 6 9-82 E No High Press SI Flow Cont Valve 2SI-VIS4681 526S 6 9-82 E No High Press Sl Flow Cont Valve 2SI-V]54OB2 527S 5 9-82 E No High Preas Si Flow Cont Valve 2SI-VI547B3 528S 6 9-82 E No High Press SI Flow Cont Valve 2SI-VI544B4 530S 4 9-2-80 E No Low Press Safety Inj Pump A 532S 3 3-9-79 E No Shutdown & Cooling Line A Bypass Valve 2SI-FM317A 533S 2 10-4-78 E No Shutdown Cooling Flow Control Valve 2SI-FM318A 535S 4 9-2-80 E No Low Press Safety Inj Pump B 537S 2 10-4-78 E No Shutdown Cooling Line B Bypass Valve SI-FM348B 538S 2 10-4-78 E No Shutdown Cooling Flow Contr Valve SI-FM349B 539S 4 9-82 E No Shutdown Line to RCS Loop I 541S 7 9-82 E No Low Press Sl Flow Contr Valve 2SI-VI549AI 542S 7 9-82 E No Low Press Sl Flow Contr Valve 2SI-VIS39BI 543S 6 9-82 E No Low Press Sl Flow Contr Valve 2SI-VI54IA2 544S 7 9-82 E No Low Press Sl Flow Contr Valve 2SI-VI543B2 545S 4 9-11-80 E No Nitrogen Cont Isol Valve 2NG-F605 & 2NG-F604 548S 8 9-82 E No Shutdown Cooling Temp 552S 6 9-82 E No SI Tak IA Isol Valve & ISI-VSO5 TKIA 553S 9 9-82 E No SI Tnk IA Nitrogen Contr Valve & Vent Valve 554S 4 9-82 E No SI Tnk 1A Leakage Drain Valve & Fill Drain Valve 555S 6 9-82 9 No SI Tak IA Vent Valve 558S 6 9-92 E No SI Tnk IB Isol Valve ISI-VI506 TKIB
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 20 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams - Safety Related (Cont'd)
B-424 559S 6 9-82 E No SI Tnk IB Nitrogen Contr Valve & Vent Va 560S 4 9-82 E No SI Tak IR Leakage Drain Valve & Fill Drain Valve 561S 6 9-82 E No SI Tank IB Vent Valve 564S 7 9-82 E No SI Tnk 2A Isol Valve ISI-VIS07 TK2A 565S 7 9-82 E No SI Tak 2A Nitrogen Cont Valve 566S 4 9-82 E No SI Tnk 2A Leakage Drain Valve & Fill Drain Valve 567S 6 9-82 E No SI Tnk 2A Vent Valve 570S 7 9-82 E No SI Tnk 28 Isol Valve ISI-V15O8 Tk2B 571S 6 9-82 E No SI Tnk 2B Nitrogen Contr Valve & Vent Valve 572S 4 9-82 E No SI Tnk 2B Leakage Drain Valve Fill Drain Valve 573S 7 9-82 E No SI Tnk 2B Vent Valve 589S 5 9-82 E No Shutdown Cool Heat Exch Instr 590S 10 9-82 E No RCS Loop #1 Shutdown Cool Isol Valve ISI-VI502 591S 11 9-82 E No RCS Loop #1 Shutdown Cool Isol Valve ISI-VI501B 592S 4 9-82 E No RCS Loop #1 Shutdown Cool Isol Valve 2SI-V326B 593S 3 9-82 E No RCS Loop #1 Hot Leg Inj Drain Valve ISI-VI578 594S 5 9-26-79 E No Hydraulic Pump Motor RCS Loop #1 SHDN Cool Isol Valve 1SI-VI502B 595S 10 9-82 E No RCS Loop #2 Shutdown Cool 1801 Valve 1SI-VI504A 596S 9 9-82 E No RCS Loop #2 Shutdown Cool Isol Valve 1SI-VI503A 597S 3 2-21-79 E No RCS Loop #2 Shutdown Cool Isol Valve ISI-V327 598S 3 9-82 E No RCS Loop #2 Hot Leg Inj Drain Valve 1SI-VI577 599S 7 9-82 E No Hydraulic Pump Motor RCS Loop #2 SHDN Cool Isol Valve 1SI-VI503A 604S 4 9-82 E No Contaiment Spray Riser Level Pump 'A' 605S 6 9-82 E No Containment Spray Pump A
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 21 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams-Safety Related (Cont'd) B-424 607S 4 9-82 E No Contaiment Spray Isolation Valve 2CS-F305A 608S 9 9-82 E No Containment Spray Pump A Disch Press & Flow & Hisc Ann Inputs 609S 5 9-82 E No Containment Spray Riser Level Pump 'B' 610S 8 9-82 E No Containment Spray Pump B 612S 5 9-82 E No Containment Spray Isolation Valve 2CS-306B 613S 8 9-82 E No Containment Spray Pump B Disch Press & Flow & Misc Ann Inputs 618S 1 9-82 E No Containment Pressure Instrumentation Isol. Valves 620S 4 9-92 E No Containment Pressure Instrumentation SH 1 621S 6 9-82 E No Containment Pressure Instrumentation SH 2 623S 12 9-82 E No Containment Pressure & SIS Sump Temp & Level Instr SH 1 624S 14 9-82 E No Containment Pressure & SIS Sump Temp & Level Instr SH 2 625S 8 9-82 E No Containment Personnel Air Lock 680S 4 9-82 E No Waste Gas Cont Isol Valves 700S 4 9-82 E No Comp Cool Wtr Pumps Comon Circuits 2WM-FI57A/D & 2WM-FI58A/B 701S 6 6-27-90 E No Comp Cool Suct Disch HDR Isol Valve 3CC-F13A/B, 3CC-FI09A/B 702S 6 6-27-80 E No Comp Cool Suct Disch HDR Isol Valve 3CC-F14A/b, 3CC-FIIOA/B 703S 2 6-18-79 E No Comp Cool Suct Disch HDR Isol Valve 3CC-F15A/B, 3CCFIIIA/B 704S 3 9-82 E No Comp Cool Suct Disch HDR Isol Valve 3CC-F16A/B, 3CC-Fll2A/B 705S 6 9-82 E No Comp Cool Wtr Pump A 707S 7 9-82 E No Comp Cool Wtr Pump A/B 709S 4 9-82 E No Comp Cool Wtr Pump B 713S 10 9-82 E No Comp Cool Wtr System Instrumentation SH2 714S 8 9-82 E No Comp Cool Wtr System Instrumentation SH3 715S 5 1-8-80 E No Comp Cool Wtr System Instrumentation SH4 716S 6 9-82 E No Comp Cool Wtr System Instrumentation SH5 731S, 5 2-18-80 E No Dry Tower A Fan No. 1 732S 5 2-18-80 E No Dry Tower A Fan No. 2 733S 5 2-18-80 R No Dry Tower A Fan No. 3 734S 5 2-18-80 E No Dry Tower A Fan No. 4 735S 5 2-18-80 E No Dry Tower A Fan No. 5 736S 6 2-18-80 E No Dry Tower A Fan No. 6 737S 5 2-18-80 E No Dry Tower A Fan No. 7 738S 5 2-18-80 E No Dry Tower A Fan No. 8 739S 5 2-18-80 E No Dry Tower A Fan No. 9 740S 6 2-18-80 E No Dry Tower A Fan No. 10
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B)
G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 22 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams-Safety Related (Cont'd)
B-424 741S 5 2-18-80 E No Dry Tower A Fan No. 11 742S 5 2-18-80 E No Dry Tower A Fan No. 12 743S 5 2-18-80 E No Dry Tower A Fan No. 13 744S 5 2-18-80 E No Dry Tower A Fan No. 14 745S 5 2-18-80 E No Dry Tower A Fan No. 15 749S 6 9-82 E No Dry Tower Isol Valve 3CC-B201A & Bypass Valve 3CC-F265A 752S 5 9-82 E No Aux Comp Coolant Pump A 753S 7 9-82 E No Aux Comp Coolant Pump Alarms & Computer Inputs 761S 6 9-82 E No Wet Tower A Fan No. 1 762S 4 9-82 E No Wet Tower A Fan No. 2 763S 4 9-82 E No Wet Tower A Fan No. 3 764S 4 9-82 E No Wet Tower A Fan No. 4 765S 5 2-7-80 E No Wet Tower A Fan No. 5 766S 4 9-82 E No Wet Tower A Fan No. 6 767S 4 9-92 E No Wet Tower A Fan No. 7 768S 7 9-82 E No Wet Tower A Fan No. 8 770S 2 9-82 E No Cool Tower A Motor Space Heaters 781S 4 2-18-80 E No Dry Tower A Fan No. 1 782S 6 9-82 E No Dry Tower A Fan No. 2 783S 5 9-82 E No Dry Tower A Fan No. 3 754S 0 E No ACCW Pump A Disch. Line Isol. Valve 3CC-B288A 784S 6 9-82 E No Dry Tower B Fan No. 4 785S 5 9-82 E No Dry Tower B Fan No. 5 786S 5 9-82 E No Dry Tower B Fan No. 6 787S 5 9-82 E No Dry Tower B Fan No. 7 788S 5 9-82 E No Dry Tower B Fan No. 8 789S 4 2-18-80 E No Dry Tower B Fan No. 9 790S 4 2-18-80 E No Dry Tower B Fan No. 10 791S 4 2-18-80 E No Dry Tower B Fan No. 11 792S 6 9-82 E No Dry Tower B Fan No. 12 793S 4 2-18-80 E No Dry Tower B Fan No. 13 794S 4 2-18-80 E No Dry Tower B Fan No. 14 795S 7 2-18-80 E No Dry Tower B Fan No. 15 799S 4 4-17-79 E No Dry Tower B Isol Valve 3CC-B203B & Bypass Valve 3CC-B262B 802S 4 9-82 E No Auxiliary Component Cooling Water Pump B 803S 6 9-82 E No Auxiliary Component Coolant Pump B Alarms and Computer Inputs 811S 4 5-16-79 E No Wet Tower B Fan No. 1
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 23 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams-Safety Related (Cont'd)
B-424 812S 4 9-82 E No Wet Tower B Fan No. 2 813S 4 9-82 E No Wet Tower B FAn No. 3 814S 4 9-82 E No Wet Tower B Fan No. 4 815S 5 9-82 E No Wet Tower B Fan No. 5 816S 5 9-82 E No Wet Tower B Fan No. 6 817S 5 9-82 E No Wet Tower B Fan No. 7 818S 6 9-82 E No Wet Tower B Fan No. 8 819S 3 1-3-79 E No Wet Tower Level Equalizing Valves Cross Tie Valves 820S 2 9-82 E No Cool Tower Motor Space Materials 821S 4 7-24-79 E No Chillers Coolant Selective Valves Common CKT System "A" 822S 2 10-9-78 E No Chillers Coolant Selective Valves 3CC-F272B & 3CC-F274A 823S 2 10-9-78 E No Chillers Coolant Selective Valves 3CC-F276A & 3CC-F278A 824S 6 9-82 E No Containment Fan Coolers Cool Coils Instr System "A" 826S 3 1-3-79 E No Chillers Coolant Selective Valves Comon CKT System "B" 827S 3 5-10-79 E No Chillers Coolant Selective Valves 3CC-F273B & 3CC-F275B 828S 3 5-11-79 E No Chillers Coolant Selective Valves 3CC-F277H & 3CC-F279B 829S 7 9-82 E No Containment Fan Coolers Cool Coils Instr Syst "B" 834S 5 6-27-80 E No Component Coolant Isol Valve 3CC-FI22A 836S 6 9-82 E No Component Coolant Isol Valve 3CC-Fl239 838S 2 9-25-78 E No Component Coolant Isol Valve 3CC-133A/B 840S 5 9-82 E No Fuel Pool Temp. Control Valve 3CC-FM 138AB CH A 841S 4 9-82 E No Fuel Pool Temp. Control Valve 3CC-FM 138AB CH B 804S 0 E No ACCW Pump B Disch. Line Isol. Valve 3CC-B289B 843S 7 3-3-80 E No Component Coolant Surge Tank Level Contr Valve 7OW-LM601 845S 6 9-27-80 E No CCW to RC Pumps Coolant Valve 2CC-FI46A/B 846S 8 9-82 E No CCW from RC Pumps Isol Valve 2CC-FI47A/B & 2CC-F243A/B 847S 3 5-11-79 E No Shutdown Heat Exch A Outlet Valve 3CC-FI30A 848S 3 5-11-79 E No Shutdown Heat Exch B Outlet Valve 3CC-FI31B 851s 8 9-82 E No Component Coolant Makeup Pump A 852S 8 9-82 E No Component Coolant Makeup Pump 8 854S 3 4-10-79 E No CCW Makeup to Return Header Valve 3CC-F240A 855S 3 4-10-79 E No CCW Makeup to Return Header Valve 3CC-F241B 859S 7 9-82 E No DG'S Stand Pipes Level Contr Valve 3CC-L640A & 3CC-L641B 862S 9 9-82 E No Containment Sump Pump Isolation Valves 908S 0 9-82 E No Toxic Chemical Detection System SH. 5
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 24 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams-Safety Related (Cont'd)
B-424 911S 7 9-82 E No Coolant Sampling Isol Valves 2SL-FI501A/B & 2SL- 1504A/B 912S 6 9-82 E No Pressurized Surge Line Cont Isol Valve 2SL-FI502 913S 7 9-82 E No PZRD STH Space Sample Cont Isol Valves 2SL-FIS03A/B & 2SL-FI506A/B 922S 8 9-82 E No STH GEN I Sample Isol Valves 2SL-F601 & 2SL-F602 923S 5 9-82 E No STH GEN 2 Sample Isol Valves 2SL-F603 & 2SLF604 929S 4 8-27-80 E No STH Line Sample Isol Valves 2MS-F714 & 2MS-F715 931S 3 9-82 E No Reactor Coolant System Vent Valves SH. 1 932S 5 9-82 E No Reactor Coolant System Vent Valves SH. 2 933S 3 9-82 E No Reactor Coolant System Vent Valves SH. 3 934S 3 9-82 E No Reactor Coolant System Vent Valves SH. 4 940S 1 9-82 E No SIS Sump Sampling Cont. Isol. Valve 2SI-E648 951S 5 9-82 E No Hydrogen Analyzer System A Sh 1 952S 8 9-82 E No Hydrogen Analyzer System A Sh 2 953S 1 11-10-78 E No Hydrogen Analyzer System A Inlet Pump 954S 1 11-10-78 E No Hydrogen Analyzer System A Outlet Pump 955S 3 9-82 E No Hydrogen Analyzer System B Sh 1 956S 6 9-82 E No Hydrogen Analyzer System B Sh 2 957S 1 11-10-78 E No Hydrogen Analyzer System B Inlet Pump 958S 2 5-11-79 E No Hydrogen Analyzer System B Outlet Pump 960S 8 9-82 E No Hydrogen Recombiner System A 962S 7 9-82 E No Hydrogen Recombiner System 8 997S 5 8-27-80 E No Instrument Air Cont Isol Valve 21A-F601AB 0(DRN 01 758, R11 A) 1000S 7 9-10-79 E No Charging Pump A Cooler AH-18 (3A-SA) 1001S 7 9-10-79 E No Charging Pump AB Cooler AH-22 (3A-SAB) 1002S 6 9-10-79 E No Charging Pump B Cooler AH-18 (3B-SB) 1003S 6 9-10-79 E No Charging Pump AB Cooler AH-22 (3B-SAB) 1004S 7 9-82 E No Safeguard Pump Room A Cooler AH-21 (3-SAB) 1005S 11 9-82 E No Safeguard Pump Room A Cooler AH-2 (3A-SA) 1006S 9 9-82 E No Safeguard Pump Room A Cooler AH-2 (3A-SA) 1007S 8 9-82 E No Safeguard pump Room A Cooler AH-2 (3B-SB) 1008S 8 9-82 E No Safeguard Pump Room A Cooler AH-2 (3D-SB)
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 25 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams-Safety Related (Cont'd) B-424 1010S 7 9-10-79 E No CCW Pump A Cooler AH-10 (3A-SA) 1011S 7 9-82 E No CCW Heat Exch A Cooler AH-24 (3A-SA) 1012S 7 9-12-79 E No CCW Pump AB Cooler AH-20 (3A-SAB) 1013S 7 9-12-79 E No CCW Pump AB Cooler AH-20 (3B-SAB) 1014S 6 9-10-79 E No CCW Pump B Cooler AH-10 (3B-SB) 1015S 7 8-25-80 E No CCW Heat Exch B Cooler AH-24 (3B-SB) 1017S 8 8-25-80 E No Shutdown Heat Exch A Cooler AH-3 (3A-SA) 1018S 8 8-25-80 E No Shutdown Heat Exch B Cooler AH-3 (3B-SB) 1020S 8 9-10-79 E No Emerg FWP A Cooler AH-17 (3A-SA) 1021S 8 9-10-79 E No Emerg FWP A Cooler AH-17 (38-SB) 1026S 8 8-25-80 E No Equipt Room Cooler AH-26 (3A-SA) 1027S 8 9-82 E No Equipt Room Cooler AH-26 (3B-SB) 1031S 5 9-82 E No Control Atmos Release System Supply Fan S-3 (3A-SA) 1032S 9 9-82 E No Control Atmos Release System Exhaust Fan E-18 (3A-SA) G(DRN 01 758, R11 A) 1033S 5 9-82 E No Control Atmos Release System 0(DRN 01 758, R11 A) Disch Valve 2KV-BI67A 1034S 7 9-82 E No Control Atmos Release System Suct Valve-F253A 1035S 5 8-92 E No Control Atmos Release System Supply Fan S-3 (38-SB) 1036S 11 9-82 E No Control Atmos Release System Exhaust Fan E-18 (3B-SB) Disch Valve 2HV-BI689 1037S 4 9-82 E No Control Atmos Release System Disch Valve 2HV-BI68B 1038S 7 9-82 E No Control Atmos Release System Shut Valve 2HV-F254B 1040S 9 9-92 E No Diesel Gen A Room Vent System
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 26 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams-Safety Related (Cont'd) B-424 1041S 5 9-82 E No Diesel Gen A Room Exhaust Fan E-28 (3A-SA) 1042S 7 9-3-80 E No Diesel Gen B Room Vent System 1043S 5 9-82 E No Diesel Gen B Room Exhaust Fan E-28 (3B-SB) 1045S 5 9-82 E No Water Chiller Compressor WC-1 (3A-SA) 1046S 5 9-82 E No Water Chiller Compressor WC-1 (3A-SA) Oil Pump 1047S 3 4-17-79 E No Water Chiller WC-1 (3A-SA) Pump Out Compressor 1048S 6 9-82 E No Chilled Water Pump P-1 (3A-SA) 1049S 9 9-82 E No Chilled Water Sys RCC Va & Exp Tnk Lev Cnt Va 1050S 7 8-25-80 E No Chilled Water System A Alarm & Computer Inputs 1051S 7 2-15-80 E No Chilled Water System A Instrumentation Sh 1 1052S 6 9-82 E No Chilled Water System A Instrumentation Sh 2 1055S 6 9-82 E No Water Chiller Compressor WC-1 (3B-SB) 1056S 4 9-82 E No Water Chiller WC-1 (3B-SB) Oil Pump 1057S 4 9-82 E No Water Chiller WC-1 (38-SB) Pumpout Compressor 1058S 7 9-26-79 E No Chilled Water Pump P-1 (3B-SB) 1059S 8 9-82 E No Chilled Water Sys B Recirc Va & Exp Tank Level Contr Va 1060S 6 9-82 E No Chilled Water Sys B Alarm & Computer Inputs 1061S 5 9-14-79 E No Chilled Water Sys B Instrumentation Sh 1 1062S 5 9-14-79 E No Chilled Water Sys 8 Instrumentation Sh 2 1065S 6 9-82 E No Water Chiller Compressor WC-1 (3C-SAB) 1066S 6 9-82 E No Water Cbiller WC-1 (3C-SAB) Oil Pump 1067S 4 9-82 E No Water Chiller WC-1 (3C-SAB) Pumpout Compressor 1068S 8 9-82 E No Chilled Water Pump P-1 (3C-SAB) 1069S 8 9-82 E No Chilled Water System AB Recirc Va & Exp Tank Level Contr Va 1070S 7 9-82 E No Chilled Water System AB Alarm & Computer Inputs 1071S 5 9-14-79 E No Chilled Water System AB Instrumentation Sh I 1072S 4 9-14-79 E No Chilled Water System AB Instrumentation Sh 2 1073S 7 9-82 E No Chilled Water System Isolation Valves Sh I 1074S 6 9-82 E No Chilled Water System Isolation Valves Sh 2 1075S 7 9-82 E No Swgr Area Air Handling Unit AH-25 (3A-SA) 1076S 11 9-82 E No AH-25 (3A-SA) Electric Heating Coil EHC-36 (3A-SA) 1077S 9 9-82 E No AH-25 (3A-SA) Chilled Water Contr Valve & Hisc 1079S 5 9-82 E No Swgr Area Air Handling Unit AH-25 (3B-SB) 1080S 10 9-82 E No AH-25 (3B-SB) Electric Heating Coil EHC-36 (3B-SB) 1081S 9 9-82 E No AH-25 (3B-SB) Chilled Water Contr Valve & Misc
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 27 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams-Safety Related (Cont'd) B-424 1083S 6 9-82 E No Battery Room A Exhaust Fan E-29 (3A-SA) 1084S 8 9-82 E No Battery Room A Exhaust Fan E-29 (3B-SB) 1085S 8 9-82 E No Battery Room Fans Cooler E-52 (3A-SA) 1086S 5 9-82 E No Battery Room AB Exhaust Fan E-31 (3A-SA) 1087S 4 11-14-79 E No Battery Room AB Exhaust Fan E-31 (3B-SB) 1088S 6 9-82 E No Battery Room Fan Cooler E-52 (3B-SB) 1089S 4 5-7-79 E No Battery Room B Exhaust Fan E-30 (3B-SB) 1090S 5 9-82 E No Battery Room Exhaust Fan E-30 (3B-SB) 1092S 5 5-7-79 E No Computer Battery Room Exhaust Fan E-00 (3A-SA) 1093S 5 11-28-79 E No Computer Battery Room Exhaust Fan E-00 (3B-SB) 1099S 7 9-82 E No Annulus Negative Press Sys Sh 3 1110S 3 6-18-79 E No CVAS Isolation Valves SH1 1111S 2 6-18-79 E No CVAS Isolation Valves Sh2 1113S 6 9-82 E No CVAS Exhaust Fan E-23 (3B-SA) 1114S 8 9-82 E No CVAS Exhaust Fan E-23 (3B-SB) 1115S 3 1-8-80 E No CVAS Electric Heating Coil EHC-48 (3A-SA) 1116S 3 9-82 E No CVAS Electric Heating Coil EHC-48 (3A-SA) 1117S 5 11-14-79 E No RAB HVAC Equip Room Supply Fan AH-13 (3A-SA) 1119S 5 3-19-79 E No RAB HVAC Equip,Room Supply Fan AH-13 (3B-SB) 1121S 5 11-14-79 E No RAB HVAC Equip Room Exhaust Fan E-41 (3A-SA) 1122S 5 11-14-79 E No RAB HVAC Equip Room Exhaust Fan E-41 1123S 7 9-82 E No RAB HVAC Equip Room Elec Htg Coil ECH-55 (3A) 1124S 7 9-82 E No RAB HVAC Equip Room Elec Htg Coil EHC-55 (3B) 1128S 10 9-82 E No Cont Purge Isol VA'S Shl, 1129S 9 9-82 E No Cont Purge Isol VA's Sh2 1130S 9 9-82 E No Cont Vacuum Relief VA 2HV-BI56A 1131S 12 9-82 E No Cont Vacuum Relief VA 2HV-BI57B 1132S 7 9-82 E No Cont Fan Cooler AH-1 (3A-SA) 1133S 5 9-82 E No Cont Fan Cooler AH-1 (3C-SA) 1134S 9 9-82 E No Cont Fan Cooler Syst A Valves 1135S 6 9-82 E No Cont Fan Cooler AH-1 (3B-SB) 1136S 5 9-82 E No Cont Fan Cooler AH-1 (3D-SB) G(DRN 01 758, R11 A) 1137S 8 9-82 E No Containment Fan Coolers System B Valves 1138S 8 9-82 E No Containment Fan Coolers Instrumentation
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 28 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams-Safety Related (Cont'd)
B-424 1146S 5 9-92 E No Contr Room Emerg Filtr Unit A Intake and Return Air Dampers Comm Cont SH 1 1147S 7 9-82 E No Contr Room Emerg Filtr Unit B Intake and Return Air Dampers Comm Cont SH 2 1148S 8 9-82 E NO Contr Room Emerg Filtr Unit S-8 (3A-SA) 1149S 11 9-82 E No S-8 (3A-SA) Electric Heating Coil EHC-49 (3A-SA) 1150S 5 9-82 E No Contr Room Filtr Unit S-8 (3B-SD) 1151S 10 9-82 E No S 8 (3B-SB) Electric Heating Coil EHC-49 (3B-SB) 1152S 5 9-82 E No Control Rm Emerg Filtr Unit Intake Valve 3HV-B196A 1153S 5 9-82 E No Control Rm Emerg Filtr Unit Intake Valve 3HV-B197A 1154S 3 9-82 E No Control Rm Emerg Filtr Unit Intake Valve 3HV-B198A 1155S 3 9-82 E No Control Rm Emerg Filtr Unit Intake Valve 3HV-B199A 1156S 5 9-82 E No Control Rm Emerg Filtr Unit Intake Valve 3HV-B201A 1157S 4 9-82 E No Control Rm Emerg Filtr Unit Intake Valve 3HV-B200B 1158S 3 9-82 E No Control Rm Emerg Filtr Unit Intake Valve 3HV-B203A 1159S 3 9-82 E No Control Rm Emerg Filtr Unit Intake Valve 3HV-B202B 1160S 7 9-82 E No Control Room Toilet Exhaust Fan E-34 (3A-SA) 1161S 8 9-82 E No Control Room Toilet Exhaust Fan E-34 (3B-SB) 1162S 2 9-82 E No Control Room Area Norm & Purge Dampers D-45, D-46, D-68 1164S 6 9-82 E No Contr Rm Confer Rm & Kitchen Exh Fan Disch Valves & Dampers 0(DRN 01 758, R11 A) 1165S 7 9-82 E No Contr Room Air Handling Unit AH-12 (3A-SA) 1166S 8 9-82 E No AH-12 (3A-SA) Electric Heating Coil EHC-34 (3AS) 1167S 6 9-82 E No Contr Room Air Handling Unit AH-12 (3B-SE) 1168S 7 9-82 E No AH-12 (3B-SB) Electric Heating Coil EHC-34 (3B-SB) 1232S 10 9-82 E No FHB Vent Sys Exhaust Fan E-21 (3A-SA) 1233S 10 9-82 E No FHB Vent Sys Exhaust Fan E-21 (3B-SH) 1234S 6 3-3-80 E No FHB Vent System Emerg Filter Unit E-35 (3A-SA) 1235S 8 9-82 E No E-35 (3A-SA) Electric Heating Coil
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 29 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams-Safety Related (Cont'd)
B-424 1236S 5 2-5-80 E No FHB Vent Sys Emerg Filtr Unit E-35 (3B-SB) 1237S 10 9-82 E No E-35 (3B-SR) Electric Heating Coil 1238S 5 9-82 E No FHB Isolation Dampers SH1 1239S 6 9-82 E No FHB Isolation Dampers SH2 1242S 4 8-25-80 E No SWGR Area Air Holding Unit AH-30 (3A-SA) 1243S 5 9-82 E No AH-30 (3A-SA) Chilled Wtr Contr VA & Hisc 1245S 3 8-25-80 E No SWGR Ares Air Handling Unit AH-30 (3B-SB) 1246S 7 9-82 E No AH-30 (3B-SB) Chilled Wtr Contr VA & Misc 1250S 7 9-82 E No Shield Bldg Vent Sys Fan E-17 (3A-SA)
1252S 4 9-82 E No Shield Bldg Vent Sys Train Inlet 2HV-BI60A 1253S 6 9-82 E No Shield Bldg Vent Sys Train Outlet 2HV-BI58A 1254S 8 9-82 E No Electric Heater Coil EHC-51 (3A-SA) 1255S 5 9-82 E No Shield Bldg. Vent Syst A Exch VA 2HV-BI622A 1256S 2 9-82 E No Shield Bldg. Vent Syst A Recirc VA 2HV-BI64A 1257S 8 9-82 E No Shield Bldg. Vent Syst A Aux Exch VA 2HV-BI73A 1259S 7 9-82 E No Shield Bldg. Vent Sys Fan E-17 (3B-SB) 1261S 4 9-82 E No Shield Bldg. Vent Sys Train Inlet 2HV-BI61B 1262S 6 9-82 E No Shield Bldg Vent Sys Train Outlet 2HV-:159B 1263S 5 9-82 E No Electric Heating Coil EHC-51 (3B-SB) 1264S 5 9-82 E No Shield Vent Syst B Exh VA 2HV-BI63B 1265S 2 4-5-79 E No Shield Bldg Vent Syst B Recirc VA 2HV - B165B 1266S 9 9-82 E No Shield Bldg Vent Syst B Aux Exh VA 2HV - B173B 1268S 3 9-82 E No CVAS A Train Inlet VA 3HV B208A 1269S 2 4-5-79 E No CVAS A Train Outlet VA 3HV B206A 1270S 8 9-82 E No CVAS A Train Instrumentation 1271S 4 9-82 E No CVAS B Train Inlet VA 3HV - B209B 1272S 2 4-5-79 E No CVAS B Train Outlet 3HV - B209B 1273S 9 9-82 E No CVAS Instrumentation 1274S 5 9-82 E No CVAS A Train Min Flow Damper D-71 (SA) 1275S 5 9-82 E No CVAS B Train Min Flow Damper D-71 (SB) 1326S 2 11-17-78 E No Fire Wtr Cont Isol VA's 2FP-F127 & 2FP-F129 1509S 2 9-82 E No Steam Gen. No. 1 Level (Wide Range) & EFW Flow 1510S 6 9-82 E No SG No 1 Feedwater Isolation VA 2FV-V823A SH1 1511S 8 9-82 E No SG No 1 Feedwater Isolation VA 2FV-823 SH2
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 30 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams-Safety Related (Cont'd)
B-424 1525S 2 9-82 E No Steam Gen. No. 2 Level (Wide Range) & EFV Flow 1526S 6 9-82 E No SG No 2 Feedwater Isolation VA 2FV-V824B SH1 1527S 9 9-82 E No SG No 2 Feedwater Isolation VA 2FW-V824B SH2 1530S 10 9-82 E No Condensate Storage Pool Inlet VA 6 CD-L362 & Instrumentation 1531S 6 9-82 E No Emerg Feedwater Pump A 1533S 6 9-82 E No Emerg Feedwater Pump B 1535S 2 11-14-79 E No Emerg FWPT Stm Shutoff VA 2MS-V611A 1536S 1 10-9-78 E No Emerg FWPT Sts Shutoff VA 2MS-V612B 1540S 8 9-82 E No Emergency FWPT Stop Valve 1541S 7 9-82 E No Emergency FWPT Gov Valve 1546S 5 9-82 E No SG No 1 Emerg FV Isol VA's 2FW-V848A and 2FV-V852A 1547S 4 9-82 E No SG No 1 Emerg FV Isol VA's 2FW-V8478 and 2FW-851B 1548S 5 9-82 E No SG No 2 EMerg FV Isol VA's 2FW-849A
1549S 5 9-82 E No SG No 2 Emerg FV Isol VA's 2FW-V850B and 2FW-V854B 1550S 8 9-82 E No Emergency Feedwater Instrumentation 1551S 1 9-82 E No SG No. 1 Emergency FW Contr. VA's 2FW-VB52A & V851B 1552S 1 9-82 E No SG No. 2 Emergency FV Contr. VA's 2FW-V853A & V854B 1560S 4 9-82 E No Steam Gen No. 1 Blowdown Cont Isol VA 2BD-F603 & 2RD-F604 1565S 6 9-82 E No Steam Gen No. 2 Blowdown Cont Isol VA 2BD-F605 & 2BD-F606 1643S 4 9-14-79 E No SG1 Main Sto Atmospheric Damp VA 2MS-PM629A 1644S 3 9-82 E No STM Line 1 Upstream Normal Drain VA 2MS-V670 1645S 4 9-82 E No STM Line 1 Upstream Emerg Drain VA 2MS-V671 1646S 6 9-82 E No STM Line 1 Isolation 2MS-V602A SH1 1647S 8 9-82 E No STM Line 1 Isolation 2MS-V602A SH2 1658S 5 9-14-79 E No SG2 Main Stm Atmospheric Dump VA 2MS-PM 6308 1659S 3 12-18-79 E No STM - Line 2 Upstream Normal Drain VA 2MS-V663 1660S 5 9-82 E No STM - Line 2 Upstream Emerg Drain VA 2MS-V664 1661S 6 9-82 E No STH - Line 2 Isolation VA 2MS-V604B SH1 1662S a 9-82 E No STM Line 2 Isolation VA 2MS-V604B SH2 2090S 1 9-82 E No Class IE Va's N2 Backup System No. 1 2091S 1 9-82 E No Class IE Va's N2 Backup System No. 2
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 31 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams-Safety Related (Cont'd)
B-424 2092S 2 9-82 E No Class IE Va's N2 Backup System No. 3 2093S 3 9-82 E No Class IE Va's N2 Backup System No. 4 2094S 1 9-82 E No Class IE Va's N2 Backup System No. 5 2095S 1 9-82 E No Class IE Va's N2 Backup System No. 6 2096S 2 9-82 E No Class IE Va's N2 Backup System No. 7 2097S 3 9-82 E No Class IE Va's N2 Backup System No. 8 2098S 1 9-82 E No Class IE Va's N2 Backup System Alarm 2301S 3 9-82 E No Diesel Gen A Air Compressor A No. 1 2302S 3 9-82 E No Diesel Gen A Air Compressor B No. 2 2304S 6 9-82 E No Diesel Gen A Jacket Water Circulating Pump 2305S 6 9-82 E No Diesel Gen A Jacket Water Heater 2306S 3 9-11-80 E No Diesel Gen A Cooling Water Shut-off Valve 2308S 5 9-82 E No Diesel Gen A Stanby Fuel Oil Booster Pump 2309S 6 9-82 E No Diesel Gen A Fuel Oil Tranfer Pump 2311S 5 9-82 E No Diesel Gen A Pre-Lube Oil Pump 2312S 3 9-82 E No Diesel Gen A Standby Lube Oil Pump 2313S 7 9-82 E No Diesel Gen A Lube Oil Heater 2314S 3 9-92 E No Diesel Gen A Space Htr 2315S 5 9-82 E No Diesel Gen A Generator Control Interface SH1 2316S 3 10-8-80 E No Diesel Gen A Generator Control Interface SH2 2317S 3 10-8-80 E No Diesel Gen A Generator Control Interface SH3 2318S 5 9-82 E No Diesel Gen A Engine Control Interface SH1 2319S 2 9-92 E No Diesel Gen A Engine Control Interface SH2 2320S 2 9-82 E No Diesel Gen A Engine Control Interface SH3 2321S 4 9-82 E No Diesel Gen A Engine Control Interface SH4 2322S 0 9-20-77 E No Diesel Gen A Engine Control Interface SH5 2323S 11 9-82 E No Diesel Gen A Engine Control Interface SH6 2324S 4 9-82 E No Diesel Gen A Engine Control Interface SH7 2326S 7 9-82 E No Diesel Gen B Computer Inputs SH2 2327S 6 9-82 E No Diesel Generator A Breaker 2328S 6 9-82 E No Diesel Generator A Comp Inputs SH3 2332S 7 9-82 E No 4.16 KV Bus 3A3 Tie to Bus 3A2-S 2334S 7 9-82 E No 4.16 KV Bus 3A3 Tie to Bus 3AB3-S 2335S 7 9-82 E No 4.16 KV Bus 3AB3-S Tie to Bus 3A3-S 2336S 1 9-82 E No 4.16 KV Bus 3A3 MTP Space Beaters 2337S 7 9-82 E No 4.16 KV Bus 3A3-S Undervoltage Relays Test 2338S 3 9-82 E No 4.16 KV Bus 3A3-1 Undervoltage Relays Test
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 32 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams-Safety Related (Cont'd)
B-424 2339S 5 9-82 E No 480V Bus 3A31-S Undervoltage Relays Test 2340S 4 9-82 E No 480V Bus 3A31-S Undervoltage Relays Test 2341S 4 9-82 E No Sequencer A SH1 2342S 1 9-82 E No Sequencer A SH2 2343S 1 8-27-80 E No Sequencer A SH3 2346S 0 9-82 E No STA Service Transformer 3A32 Feeder Interlocks 2347S 5 9-82 E No STA Service Transformer 3A31-S Feeder 2348S 0 9-82 E No STA Service Transformer 3A32 Feeder 2349S 5 9-82 E No STA Service Transformer 3A315S Feeder 2351S 3 9-82 E No Diesel Gen B Air Compressor No. 1 2352S 3 9-82 E No Diesel Gen B Air Compressor No. 2 2354S 6 9-82 E No Diesel Gen B Jacket Water Circulating Pump 2355S 6 9-82 E No Diesel Gen B Jacket Water Heater 2356S 3 9-82 E No Diesel Gen B Cool Water Shutoff Valve 2358S 5 9-82 E No Diesel Gen B Standby Fuel Oil Booster Pump 2359S 6 9-82 E No Diesel Gen B Fuel Oil Transfer Pump 2361S 5 9-82 E No Diesel Gen B Pre-Lube Oil Pump 2362S 3 9-82 E No Diesel Gen B Stanby Lube Oil Pump 2363S 5 9-82 E No Diesel Gen B Lube Oil Heater 2364S 2 9-82 E No Diesel Gen B Space Htr 2365S 5 9-82 E No Diesel Gen B Generator Control Interface SH1 2366S 4 10-27-80 E No Diesel Gen B Generator Control Interface SH2 2367S 4 10-27-80 E No Diesel Gen B Generator Control Interface SH3 2368S 6 9-82 E No Diesel Gen B Engine Control Interface SH1 2369S 2 9-82 E No Diesel Gen B Engine Control Interface SH2 2370S 2 9-82 E No Diesel Gen B Engine Control Interface SH3 2371S 2 9-82 E No Diesel Gen B Engine Control Interface SH4 2372S 0 9-20-77 E No Diesel Gen B Engine Control Interface SH5 2373S 10 9-82 E No Diesel Gen B Engine Control Interface SH6 2374S 4 9-82 E No Diesel Gen B Engine Control Interface SH7 2375S 5 9-82 E No Diesel Gen B Computer Inputs SH1 2376S 8 9-82 E No Diesel Gen B Computer Inputs SH2 2377S 6 9-82 E No Diesel Generator B Breaker 2378S 4 9-82 E No Diesel Generator B Computer Inputs SH3 2382S 6 9-82 E No 4.16 KV Bus 3B3-S Tie to Bus 3B2 2384S 7 9-82 E No 4.16 KV Bus 3B3-S Tie to Bus 3AB3-S
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 33 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams-Safety Related (Cont'd)
B-424 2385S 8 9-82 E No 4.16 KV Bus 3AB3-S Tie to Bus 3B3-S 2386S 3 6-4-79 E No 4.16 KB Bus 3B3-S Mtr Space Heaters 2387S 5 9-82 E No 4.16 KV Bus 3B3-S Undervoltage Relays 2388S 3 9-82 E No 4.16 KV Bus 3B3-S Undervoltage Relays 2389S 9 9-82 E No 480V Bus 3B31-S Undervoltage Relays 2390S 5 9-82 E No 480V Bus 3B31-S Undervoltage Relays Test 2391S 4 9-82 E No Sequencer B SH1 2392S 2 9-82 E No Sequencer B SH2 2393S 3 8-27-80 E No Sequencer B SH3 2396S 0 9-82 E No STA Service Transformer 3B22 Feeder Interlocks 2397S 4 9-82 E No STA Service Transformer 3B31-S Feeder 2398S 7 9-82 E No STA Service Transformer 3B32 Feeder 2399S 5 9-82 E No STA Service Transformer 3B315-S Feeder 2406S 1 7-27-78 E No 4.16 KV Bus 3AB3-S Mtr Space Heaters 2407S 2 9-82 E No 4.16 KV Bus 3AB3-S Undervoltage Relays SH1 2408S 5 6-4-79 E No 4.16 KV Bus 3AB-S Undervoltage Relays SH2 2409S 4 9-82 E No 4.16 KV Bus 3AB3-S Undervoltage Relays Test 2410S 1 3-9-79 E No 490V Bus 3AB31-S Undervoltage Relays 2411S 3 12-18-79 E No 480V Bus 3AB31-S Undervoltage Relays Test 2477S 4 9-82 E No 480V Bus 3A31-S Tie To Bus 3AB31-S 2479S 4 9-82 E No 480V MCC 3A311-S Feeder 2480S 4 9-82 E No 480V MCC 3A312-S Feeder 2481S 4 9-82 E No 480V MCC 3A313-S Feeder 2482S 4 9-82 E No 480V MCC 3A314-S Feeder 2483S 4 9-82 E No 480V MCC 3A317-S Feeder 2486S 4 9-82 E No 480V MCC 3A312-S Bus Isolation Bkr 2487S 3 9-82 E No 480V MCC 3A313-S Bus Isolation Bkr 2488S 3 9-82 E No 480V MCC 3A314-S Bus Isolation Bkr 2489S 3 9-82 E No Computer Primary Feeder 2502S 5 9-82 E No 480V Bus 3B31-S Tie To Bus 3AB31S 2504S 4 9-82 E No 480V MCC 3B311-S Feeder 2505S 4 9-82 E No 480V MCC 3B312-S Feeder 2506S 4 9-82 E No 480V MCC 3B313-S Feeder 2507S 4 9-82 E No 480V MCC 3B314-S Feeder 2508S 4 9-82 E No 480V MCC 3B317-S Feeder 2511S 3 9-82 E No 480V MCC 3B312-S Bus Isolation Bkr 2512S 4 9-82 E No 480V MCC 3B313-S Bus Isolation Bkr 2513S 3 9-82 E No 480V MCC 3B314-S Bus Isolation Bkr 2525S 3 9-10-79 E No 480V Bus 3AB31-S Undervoltage Relays SH2
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 34 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams-Safety Related (Cont'd)
B-424 2527S 3 9-82 E No 480V Bus 3AB31-S Tie to Bus 3A31-S 2529S 3 9-82 E No 480V Bus 3AB31-S Tie to Bus 3B31-S 2531S 4 9-82 E No 480V MCC 3AB311-S Feeder 2532S 4 9-82 E No 480V MCC 3AB312-S Feeder 2533S 4 9-82 E No 480V MCC 3AB313-S Feeder 2534S 1 9-82 E No Computer Secondary Feeder 2537S 1 9-3-80 E No 480V MCC 3A315-S Bus Boltage 2538S 1 9-82 E No 208/120V AC PWR PNL 3B1-SA Feeder 2541S 1 9-3-80 E No 480V MCC 3B315-S Bus Voltage 2542S 1 11-28-79 E No 208/120V AC PWR PHL 3BI-SB Feeder 2550S 3 11-28-79 E No Battery Charger 3A1-S Feeder 2551S 3 11-28-79 E No Battery Charger 3A2-S Feeder 2552S 3 11-28-79 E No Battery Charger 3B1-S Feeder 2553S 3 11-28-79 E No Battery Charger 3B2-S Feeder 2554S 3 9-82 E No Battery Charger 3AB-1S Feeder 2555S 3 11-28-79 E No Battery Charger 3AB-2S Feeder 2560S 3 9-82 E No 125V DC Battery Charger 3A1-S and 3A2-S 2561S 4 9-82 E No 125V DC Battery Charger 3AB-S and 3AB2-S 2562S 3 9-92 E No 125V DC Battery Charger 3B1-S and 3B2-S 2563S 13 9-82 E No 125V DC Buses Instrumentation 2564S 4 9-18-80 E No Static Uninterrupt Pwr Supp 3MA-S N Feeeder 2565S 4 9-18-80 E No Static uninterrupt Pwr Supp 3MB-S N Feeder 2566S 4 9-18-80 E No Static Uninterrupt Pwr Supp 3MC-S N Feeder 2567S 4 9-18-80 E No Static Uninterrupt Pwr Supp 3ND-S N Feeder 2568S 3 11-28-79 E No Static Uninterrupt Pwr Supp 3A-S N Feeder 2569S 3 11-28-79 E No Static Uninterrupt Pwr Supp 3B-S N Feeder 2570S 7 9-82 E No Static Uninterrupt Pwr Supp 3MA-S and 3MB-S 2571S 7 8-82 E No Static Uninterrupt Pwr Supp 3MC-S and 3MD-S 2573S 4 8-6-80 E No Static Uninterrupt Pwr Supp 3A-S and 3B-S 2575S 4 9-18-80 E No Static Uninterrupt Pwr Supp 3KA-S Alt Feeder 2576S 4 9-18-90 E No Static Uninterrupt Pwr Supp 3MB-S Alt Feeder 2577S 4 9-18-80 E No Static Uninterrupt Pwr Supp 3MC-S Alt Feeder 2578S 4 9-18-80 E No Static Uninterrupt Pwr Supp 3ND-S Alt Feeder 2583S 3 11-28-79 E No Static Uninterrupt Pwr Supp 3A-S Alt Feeder 2584S 3 9-82 E No Static Uninterrupt Pwr Supp 3B-S Alt Feeder 2596S 2 9-82 E No Radiation Monitoring Safety Loop "A" 2597S 0 9-82 E No Radiation Monitors Safety Ch. A Annunciation 2598S 2 9-82 E No Radiation Monitoring Safety Loop "B" 2599S 0 9-82 E No Radiation Monitors Safety Ch. B Annunciation
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 35 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams-Safety Related (Cont'd)
B-424 2635S 0 9-82 E No Containment Atmosphere High Range Rad. Monitor A 2636S 0 9-82 E No Containment Atmosphere High Range Rad. Monitor B 2656S 3 9-82 E No Component Cooling Water Process Radiation Monitor "A" 2657S 3 9-82 E No Component Cooling Water Process Radiation Monitor "B" 2660S 4 9-82 E No Containment Purge Isolation "A" Area Radiation Monitor 2661S 6 9-82 E No Containment Purge Isolation "B" Area Radiation Monitor 2662S 4 9-82 E No Containment Purge Isolation "A" Area Radiation Monitor 2663S 4 9-82 E No Containment Purge Isolation "B" Area Radiation Monitor 2666S 2 9-82 E No Shield Building EL 46.0'-140º Post-LOCA Radiation Monitor "B" 2667S 2 9-82 E No Shield Building EL 21.0'-230º Post-LOCA Radiation Monitor "B" 2668S 2 9-82 E No Shield Building EL 21.0-140º Post-LOCA Radiation Monitor "A" 2669S 2 9-82 E No Shield Building EL 46.0'-240º Post-LOCA Radiation Monitor "A" 2670S 3 9-82 E No Containment Atmosphere Airborne Radiation Monitor 2671S 6 9-82 E No Plant Stack Airborne Radiation Monitor "A" 2672S 6 9-82 E No Plant Stack Airborne Radiation Monitor "B" 2673S 6 9-82 E No Containment Atmosphere Radiation Monitor Isol. Valves 2675S 5 9-82 E No Control Room Isolation "A" Airborne Radiation Monitor 2679S 1 10-23-80 E No Control Room Isolation "B" Airborne Radiation Monitor 2681S 2 9-82 E No Control Room Isolation "B" Airborne Radiation Monitor 2685S 3 9-82 E No F/H Building Isolation "A" Airborne Radiation Monitor 2686S 2 9-82 E No F/H Building Isolation "A" Airborne Radiation Monitor 2687S 2 9-82 E No F/H Building Isolation "B" Airborne Radiation Monitor 2688S 3 9-82 E No F/H Building Isolation "B" Airborne Radiation Monitor 2694S 3 9-82 E No RPS Trip Alarms SH. 1 2695S 2 9-82 E No RPS Trip Alarms SH. 2 2696S 2 9-82 E No RPS Trip Alarms SH. 3 2697S 3 9-82 E No RPS Pretrip Alarms SH. 1 2698S 3 9-82 E No RPS Pretrip Alarms SH. 2 2699S 3 9-82 E No RPS Pretrip Alarms SH. 3 2700S 2 8-1-79 E No CP-10 External Connections SH1 2701S 3 8-1-79 E No CP-10 External Connections SH2 2702S 3 8-1-79 E No CP-10 External Connections SH3 2703S 3 8-1-79 E No CP-10 External Connections SH4 2704S 2 9-7-79 E No CP-10 External Connections SH5 2705S 3 9-82 E No CP-10 External Connections SH6 2706S 2 9-7-79 E No CP-10 External Connections SH7 2707S 2 9-7-79 E No CP-10 External Connections SH8 2708S 4 9-82 E No CP-10 External Connections SH9
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 36 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams-Safety Related (Cont'd)
B-424 2709S 4 9-82 E No CP-10 External Connections SH10 2710S 2 9-7-79 E No CP-10 External Connections SH11 2711S 2 9-7-79 E No CP-10 External Connections SH12 2712S 4 9-82 E No CP-10 External Connections SH13 2713S 5 9-82 E No CP-10 External Connections SH14 2714S 5 9-82 E No CP-10 External Connections SH15 2716S 3 9-82 E No CP-10 External Connections SH17 2717S 3 9-82 E No CP-10 External Connections SH18 2718S 3 9-92 E No CP-10 External Connections SH19 2719S 3 9-82 E No CP-10 External Connections SH20 2720S 3 9-82 E No CP-10 External Connections SH21 2721S 3 11-17-78 E No CP-10 External Connections SH22 2722S 3 11-17-78 E No CP-10 External Connections SH23 2723S 3 11-17-78 E No CP-10 External Connections SH24 2724S 5 9-82 E No Plant Protection System Renote Control Module SH1 2725S 3 9-7-79 E No Plant Protection System Remote Control Module SH2 2726S 4 9-82 E No Plant Protection System Renote Control Module SH3 2727S 3 9-7-79 E No Plant Protection System Remote Control Module SH4 2934S 1 9-82 E No Annunciator Display Cabinets SA &SB 2935S 6 9-82 E No Annunciator Logic Panel Safety Channel A and B 2951S 0 9-82 E No Reactor Coolant Flow Monitor SH. 1 2952S 0 9-82 E No Reactor Coolant Flow Monitor SH. 2 2954S 0 9-82 E No Reactor Coolant Level Monitor No. 1 SH. 1 2955S 0 9-82 E No Reactor Coolant Level Monitor No. 1 SH. 2 2956S 0 9-82 E No Reactor Coolant Level Monitor No. 2 SH. 1 2957S 0 9-82 E No Reactor Coolant Level Monitor No. 2 SH. 2 2961S 0 9-82 E No QSPDS Miscellaneous Inputs SH. 1 2962S 0 9-82 E No QSPDS Miscellaneous Inputs SH. 2 2963S 0 9-82 E No QSPDS Miscellaneous Inputs SH. 3 2964S 0 9-82 E No QSPDS Miscellaneous Inputs SH. 4 2965S 0 9-82 E No QSPDS Miscellaneous Inputs SH. 5 2966S 0 9-82 E No QSPDS Miscellaneous Inputs SH. 6 2967S 0 9-82 E No QSPDS Miscellaneous Inputs SH. 7 2968S 0 9-82 E No QSPDS Miscellaneous Inputs SH. 8 2969S 0 9-82 E No QSPDS Miscellaneous Inputs SH. 9 3001S 1 9-82 E No ESFAS Test Module Channel A SH. 1 3002S 1 9-82 E No ESFAS Test Module Channel A SH. 2 3003S 1 9-82 E No ESFAS Test Module Channel A SH. 3 3004S 1 9-82 E No ESFAS Test Module Channel A SH. 4
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 37 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams-Safety Related (Cont'd)
B-424 3005S 1 9-82 E No ESFAS Test Module Channel A SH. 5 3006S 1 9-82 E No ESFAS Test Module Channel A SH. 6 3007S 1 9-82 E No ESFAS Test Module Channel A SH. 7 3008S 1 9-82 E No ESFAS Test Module Channel A SH. 8 3009S 1 9-82 E No ESFAS Test Module Channel A SH. 9 3010S 1 9-82 E No ESFAS Test Module Channel A SH. 10 3011S 1 9-82 E No ESFAS Test Module Channel A SH. 11 3012S 1 9-82 E No ESFAS Test Module Channel A sH. 12 3013S 1 9-82 E No ESFAS Test Module Channel A SH. 13 3014S 1 9-82 E No ESFAS Test Module Channel A SH. 14 3015S 1 9-82 E No ESFAS Test Module Channel A SH. 15 3016S 1 9-82 E No ESFAS Test Module Channel A SH. 16 3017S 1 9-82 E No ESFAS Test Module Channel A SH. 17 3018S 1 9-82 E No ESFAS Test Module Channel A SH. 18 3019S 1 9-82 E No ESFAS Test Module Channel A SH. 19 3031S 1 9-82 E No ESFAS Test Module Channel B SH. 1 3032S 1 9-82 E No ESFAS Test Module Channel B SH. 2 3033S 1 9-82 E No ESFAS Test Module Channel B SH. 3 3034S 1 9-82 E No ESFAS Test Module Channel B SH. 4 3035S 1 9-82 E No ESFAS Test Module Channel B SH. 5 3036S 1 9-82 E No ESFAS Test Module Channel B SH. 6 3037S 1 9-82 E No ESFAS Test Module Channel B SH. 7 3038S 1 9-82 E No ESFAS Test Module Channel B SH. 8 3039S 1 9-82 E No ESFAS Test Module Channel B SH. 9 3040S 1 9-82 E No ESFAS Test Module Channel B SH. 10 3041S 1 9-82 E No ESFAS Test Module Channel B SH. 11 3042S 1 9-82 E No ESFAS Test Module Channel B SH. 12 3043S 1 9-82 E No ESFAS Test Module Channel B SH. 13 3044S 1 9-82 E No ESFAS Test Module Channel B SH. 14 3045S 1 9-82 E No ESFAS Test Module Channel B SH. 15 3046S 1 9-82 E No ESFAS Test Module Channel B SH. 16 3047S 1 9-82 E No ESFAS Test Module Channel B SH. 17 3048S 1 9-82 E No ESFAS Test Module Channel B SH. 18 3049S 1 9-82 E No ESFAS Test Module Channel B SH. 19
(C) Control Wiring Diagrams - Non-Safety Related
117 5 9-82 E No Excore Neutron Flux Control Channel #1 and Startup Channel #1 118 5 9-82 E No Excore Neutron Flux Control Channel #2 and Startup Channel #2
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 38 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams - Non-Safety Related (Cont'd)
190 6 9-82 E No Control Board Mounted Nuclear Instrumentation H4 192 3 9-82 E No Control Board Miscellaneous Instrumentation SH1 274 8 9-82 E No Pressurizer Level SH2 330 5 1-9-80 E No Boric Acid Batching Tank Heater 990 6 9-82 E No Instrument Air Compressor A 993 7 9-82 E No Instrument Air Compressor B Alarms & Computer Inputs 995 7 9-82 E No Instrument Air Dryer 755 0 E No ACCW Jockey Pump A 805 0 E No ACCW Jockey Pump B 1029 10 9-82 E No Reactor Cavity Cool Syst. Fan S-2(3B) 1105 6 9-82 E No RAB Normal Exhaust Fan E-22(3A) 1228 7 9-82 E No FHB Vent System Air Handling Unit AH-14(3) 1230 7 9-82 E No FHB Vent System Exh. Fan E-20(3A) 1231 7 9-82 E No FHB Vent System Exh. Fan E-20(3B) 1249 4 9-82 E No Cable Vault Area Dampers 1750 5 11-28-79 E No Turbine Trip SH 1 1751 4 11-28-79 E No Turbine Trip SH 2 1752 9 9-82 E No Turbine Trip SH 3 1753 5 9-82 E No Turbine Trip SH 4 1754 8 9-82 E No Turbine Trip SH 5 1755 5 9-82 E No Turbine Trip SH 6 1756 1 1-23-79 E No Auto Stop Reset & Vacuum Trip Latch 1757 5 9-7-79 E No Turbine Trip Alarms & Computer Inputs SH1 1758 11 9-82 E No Turbine Trip Alarms & Computer Inputs SH2 2103 4 2-15-80 E No Exitation System Annunciation 2151 7 9-82 E No Generator Metering Current SH1 2152 2 9-10-79 E No Generator Metering Current SH2 2155 3 3-28-80 E No Generator Potential Tranof Circuits 2156 3 4-5-79 E No Generator Metering Potential Circuits 2161 3 3-28-80 E No Generator Diff. & Ground Relaying 2165 1 7-5-78 E No Gen.-Main Transf. Diff. Relay SH1 2166 1 7-5-78 E No Gen.-Main Transf. Diff. SH2 2168 2 8-1-79 E No Main Transformer Gen. Breaker Diff. Relay 2184 1 7-25-78 E No Main Transf. 3A Differential Relay 2189 1 6-29-78 E No Main Transf. 3B Differential Relay 2194 1 6-29-78 E No Unit Aux Transf. 3A Differential Relay 2199 1 6-29-78 E No Unit Aux Transf. 3B Differential Relay 2201 7 9-82 E No Generator Lockout Relay 86G1 Sheet 1 2202 8 9-82 E No Generator Lockout Relay 86G1 Sheet 2
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 39 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams - Non-Safety Related (Cont'd)
2203 8 9-82 E No Generator Lockout Relay 86G1 Sheet 3 2205 8 9-82 E No Generator Lockout Relay 86G2 Sheet 1 2206 4 9-10-79 E No Generator Lockout Relay 86G2 Sheet 2 2207 6 9-82 E No Generator Lockout Relay 86G2 Sheet 3 2209 6 9-82 E No Distance Lockout Relay 2210 8 9-82 E No Synchronizing SH1 2211 3 9-82 E No Synchronizing SH2 2221 2 9-82 E No Generator Circuit Breaker 3A SH1 2222 3 9-82 E No Generator Circuit Breaker 3A SH2 2225 2 9-82 E No Generator Circuit Breaker 3B SH1 2226 2 9-82 E No Generator Circuit Breaker 3B SH2 2231 5 9-10-79 E No Aux. Transf. 3A to Bus 3A1 Breaker 2233 9 9-82 E No Aux. Transf. 3A to Bus 3A2 Breaker 2235 6 9-82 E No Aux. Transf. 3B to Bus 3B1 Breaker 2237 6 8-18-80 E No Aux. Tranef. 3B to Bus 3B2 Breaker 2238 2 12-22-78 E No 4 kV/7kV Transfer Failure Alarm 2240 3 9-82 E No Standby Transf. 3A Disconnect SW 2244 1 4-12-78 E No Standby Transf. 3A Diff Relays 2245 6 9-82 E No Standby Transf. 3A Lockout Relay 2246 10 9-82 E No Standby Transf. 3A to Bus 3A1 Breaker 2248 12 9-82 E No Control Room North Wall Area Radiation Monitor 2601 2 9-82 E No HVAC Floor EL 46.0' Radiation Monitor 2602 2 9-82 E No Reactor Aux Bldg EL 21.0' Area Radiation Monitor 2603 2 9-82 E No Reactor Aux Bldg EL-4.0' Area Radiation Monitor 2604 2 9-82 E No Reactor Aux Bldg. EL-4.0' Area Radiation Monitor 2605 2 9-82 E No Counting Room Area Radiation monitor 2606 2 9-82 E No Sample Room Ares Radiation Monitor 2607 2 9-82 E No Boric Acid Preconcentrators Filter Area Radiation Monitor 2608 2 9-82 E No Waste Filter Wall Area Radiation Monitor 2609 2 9-82 E No Druming Station Ares Radiation Monitor 2610 2 9-82 E No Charging Pumps Wall Area Radiation Monitor 2611 2 9-82 E No Reactor Aux. Bldg. Elevator Shaft Area Radiation Monitor G(DRN 01 758, R11 A) 2612 2 9-82 E No Radio Chemistry Lab Area Radiation lionitor 0(DRN 01 758, R11 A) 2613 2 9-82 E No Reactor Aux. Bldg Valve Gallery Area Radiation Monitor 2614 2 9-82 E No Reactor Aux. Bldg. Filter Flush Area Radiation Monitor 2615 2 9-82 E No Standby Transf. 3A to Bus 3A2 Breaker 2250 4 9-82 E No Standby Transf. 3B Disconnect SW
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 40 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams - Non-Safety Related (Cont'd)
2254 1 7-5-78 E No Standby Transf. 3B Diff Relays 2255 7 9-82 E No Standby Transf. 3B Lockout Relay 2256 9 9-82 E No Standby Transf. 3B to 3B1 Breaker 2258 11 9-82 E No Standby Transf. 3B to Bus 3B2 Breaker 2260 4 6-6-79 E No Main & Standby Transf 3A & 3B Ground Protection 2325 7 9-82 E No Diesel Gen A Computer Inputs SH1 2331 4 9-82 E No 4.16 kV Bus 3A2 Tie to Bus 3A3-S 2380 2 2-4-80 E No Diesel Gen B Local Annunciator Front View 2381 6 9-82 E No 4.16 kV Bus 382 Tie to Bus 3B3-S 2590 3 9-82 E No Radiation Monitor NIS Loop "A" 2592 2 9-82 E No Radiation Monitor NIS Loop "B" 2594 3 9-82 E No Radiation Monitor N/S Loop "C" 2600 2 9-82 E No RAB Waste Gas Coup Area Radiation Monitor 2616 2 9-82 E No RAB EL-35.0' Area Radiation Monitor 2617 2 9-82 E No RAB EL-4.0' Area Radiation Monitor 2618 2 9-82 E No Decontamination Room Area Radiation Monitor 2619 2 9-82 E No Gas Decay Tanks Area Radiation Monitor 2625 2 9-82 E No Spent Fuel Storage Pool Area Radiation Monitor 2626 2 9-82 E No New Fuel Vault Area Radiation Monitor 2627 2 9-82 E No Fuel Pool Pumps Area Radiation Monitor 2630 3 9-82 E No Refueling Machine Area Radiation Monitor 2631 3 9-82 E No Containment Southwest Staircase Area Radiation Monitor 2632 3 9-82 E No Containment Northeast Staircase Area Radiation Monitor 2633 3 9-82 E No Personnel Lock Annulus Area Radiation Monitor 2640 3 9-82 E No Fuel Handling Exhaust A Radiation Monitor 2641 3 9-82 E No Fuel Handling Exhaust B Radiation Monitor 2642 3 9-82 E No RAB Airborne Radiation Monitor A 2643 3 9-82 E No RAB Airborne Radiation Monitor B 2644 3 9-82 E No RAB Airborne Radiation Monitor C 2645 3 9-82 E No RAB Airborne Radiation Monitor D 2650 2 9-82 E No Condenser Vacuum Pumps Process Radiation Monitor 2651 2 9-82 E No Dry Cooling Tower Sump #1 Process Radiation Monitor 2652 2 9-82 E No Dry Cooling Tower Sump #2 Process Radiation Monitor 2653 2 9-82 E No Reactor Bldg Sump Process Radiation Monitor 2654 2 9-82 E No Industrial Waste Sump Process Radiation Monitor 2655 2 9-82 E No Circulating Water Discharge Process Radiation Monitor 2658 1 9-82 E No Component Cooling Water Process Radiation Monitor A/B 2852 3 1-9-80 E No Computer Inputs SH.53
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 41 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams - Non-Safety Related (Cont'd)
B-424 2922 1 9-82 E No Annunciator Display Cabinet A 2923 2 9-82 E No Annunciator Display Cabinet B 2924 2 9-82 E No Annunciator Display Cabinet C 2925 2 9-82 E No Annunciator Display Cabinet D 2926 3 9-82 E No Annunciator Display Cabinet E 2927 2 9-82 E No Annunciator Display Cabinet F 2928 2 9-82 E No Annunciator Display Cabinet G 2929 1 9-82 E No Annunciator Display Cabinet H 2930 1 9-82 E No Annunciator Display Cabinet K 2931 2 9-82 E No Annunciator Display Cabinet L 2932 2 9-82 E No Annunciator Display Cabinet M 2933 2 9-82 E No Annunciator Display Cabinet N G246 S01 7 4-82 E No Reactor Containment Building Sect. and Details S02 3 9-78 E No Reactor Containment Building Sect. and Details S03 1 9-78 E No Reactor Containment Building Sect. and Details S04 0 4-80 E No Reactor Containment Building Sect. and Details G322 S01 7 12-78 E No Cable Vault and Elec. Equip Room Sect. and Details S02 7 11-79 E No Cable Vault and Elec. Equip Room Sect. and Details S03 4 12-78 E No Cable Vault and Elec. Equip Room Sect. and Details S04 4 12-78 E No Cable Vault and Elec. Equip Room Sect. and Details S05 2 05-80 E No Cable Vault and Elec. Equip Room Sect. and Details S06 2 12-79 E No Cable Vault and Elec. Equip Room Sect. and Details S07 2 12-79 E No Cable Vault and Elec. Equip Room Sect. and Details S08 3 05-80 E No Cable Vault and Elec. Equip Room Sect. and Details G(DRN 01 758, R11 A) S09 2 06-82 E No Cable Vault and Elec. Equip Room Sect. and Details 0(DRN 01 758, R11 A) S10 0 04-80 E No Cable Vault and Elec. Equip Room Sect. and Details
G285 3 12-78 E No Main One Line Diagram
G286 3 7-12-78 E No Key Auxiliary One Line Diagram
G287 2 11-30-78 E No 125V DC & 120V AC One Line Diagram
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 42 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams - Non-Safety Related (Cont'd)
G310 S01 1 9-5-78 E No Reactor Containment Building EL-4 Conduit, Trays and Grounding S02 2 9-5-78 E No Reactor Containment Building EL-4 Conduit, Trays and Grounding S03 2 12-79 E No Reactor Containment Building EL-4 Conduit, Trays and Grounding S04 2 06-79 E No Reactor Containment Building EL-4 Conduit, Trays and Grounding
G311 S01 7 06-79 E No Reactor Containment Building EL+21 Conduit, Trays and Grounding S02 0 6-1-79 E No Reactor Containment Building EL+21 Conduit, Trays and Grounding
G312 6 6-26-79 E No Reactor Containment Building EL+4 Conduit Trays and Grounding
G313 7 6-79 E No Electrical Penetration Area-Elevation & Details
G314 7 10-80 E No Reactor Containment Building-Sections & Details Sh1
G315 5 9-8-78 E No Reactor Containment Building-Sections & Details SH2
B316 96 Sheets 4(a) Open E No Electrical Penetrations Details
G317 S01 8 9-20-78 No Reactor Aux. Bldg. EL. +7.00 Conduit, Trays & Grounding SH1 G317 S02 10 05-82 E No Reactor Aux. Bldg. EL. +7.00 Conduit, Trays & Grounding SH2
G318 9 05-82 E No Control Room Arrangement, Conduit & Grounding
G319 S01 8 6-14-79 E No Cable Vault Conduit, Trays & Grounding SH1 G319 S02 8 6-14-79 E No Cable Vault Conduit, Trays & Grounding SH2
G320 S01 7 11-29-78 E No Electrical Equipment Room Arrangement - Grounding SH1 G320 S02 8 11-79 E No Electrical Equipment Room Arrangement - Grounding SH2
G321 S0l 9 07-80 E No Electrical Equipment Room - Conduit & Trays SH1 G321 S02 9 05-81 E No Electrical Equipment Room - Conduit & Trays SH2
G322 S0l 7 12-29-78 E No Cable Vault & Electrical Equipment Room Sections & Details - SH1 G322 S02 7 11-79 E No Cable Vault & Electrical Equipment Room Grounding SH2 G322 S03 4 12-29-78 E No Cable Vault & Electrical Equipment Room Section & Details SH3 G322 S04 4 12-29-78 E No Cable Vault & Electrical Equipment Room Section & Details SH4 G322 S05 2 05-80 E No Cable Vault & Electrical Equipment Room Section & Details SH5
Note: (a) Revision number is that of index.
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 43 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE (C) Control Wiring Diagrams - Non-Safety Related (Cont'd)
G(DRN 01 758, R11 A) G322 S06 2 12-79 E No Cable Vault & Electrical Equipment Room Section & Details SH6 0(DRN 01 758, R11 A)
G322 S07 2 12-79 E No Cable Vault & Electrical Equipment Room Section & Details SH7 G322 S08 3 12-79 E No Cable Vault & Electrical Equipment Room Section & Details SH8 G322 S09 1 10-9-78 E No
G240 S01 4 11-30-37 E No RAB El +21 Conduit, Trays & Grounding SH1 G240 S02 5 6-14-79 E No RAB El +21 Conduit, Trays & Grounding SH2
G241 S01 5 04-82 E No RAB El +46 to +81 Conduit & Grounding SH1 G241 S02 3 10-4-78 E No RAB El +46 to +81 Conduit & Grounding SH2
G242 S01 5 07-80 E No H&V Room El +46 Conduit, Trays & Grounding SH1 G242 S02 5 04-82 E No H&V Room El +46 Conduit, Trays & Grounding SH2
G324 7 12-19-78 E No Reactor Auxiliary Building EL+21 Conduit, Trays & Grounding SH2 G327 7 11-30-78 E No Reactor Auxiliary Building EL-4 Conduit, Trays & Grounding SH1 G328 7 11-30-78 E No Reactor Auxiliary Building EL-4 Conduit, Trays & Grounding SH2 G329 6 11-30-78 E No Reactor Auxiliary Building EL-4 Conduit, Trays & Grounding SH4 G330 8 05-80 E No Reactor Auxiliary Building EL-4 Conduit, Trays & Grounding SH3
G331 10 04-80 E No Electrical Penetration Area Conduit, Trays & Grounding SH3 G332 to 6-27-79 E No Electrical Penetration Area Conduit, Trays & Grounding SH2 G333 9 6-14-79 E No Reactor Auxiliary Building EL-35 Conduit, Trays & Grounding SH1 G334 9 11-79 E No Reactor Auxiliary Building EL-35 Conduit, Trays & Grounding SH2 G335 S01 9 6-26-79 E No Reactor Auxiliary Building EL-35 Conduit, Trays & Grounding SH3 G336 9 6-26-79 E No Reactor Auxiliary Building EL-35 Conduit, Trays & Grounding SH4 G337 8 12-28-79 E No Reactor Auxiliary Building Piping Penetration Area Conduit, Trays & Grounding SH1 G338 9 04-80 E No Reactor Auxiliary Building Piping Penetration Area Conduit, Trays & Grounding SH2 G339S01 5 6-15-77 E No Reactor Auxiliary Building Sections & Details SH1 G339S02 3 11-30-78 E No Reactor Auxiliary Building Sections & Details SH2
G342 8 6-14-79 E No Fuel Handling Building EL 1 & -35 Conduit, Trays & Grounding
G343 8 05-82 E No Fuel Handling Building EL+21 & 46 Conduit, Trays & Grounding G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 44 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS DRAWING SHEET REVISION PREPARED PROPRIETARY NO. NO. NO. DATE BY INFO. TITLE
G344 S01 5 9-14-78 E No Electrical Equipment Room Switchgear, MCC & Panel Details - SH1 S02 5 04-82 E No Electrical Equipment Room Switchgear, MCC & Panel Details - SH2 S03 3 9-14-78 E No Electrical Equipment Room Switchgear, MCC & Panel Details - SH3
G373S01 8 11-30-78 E No Cooling Tower Area - Conduit & Grounding-SH1 S02 7 11-30-78 E No Cooling Tower Area - Conduit & Grounding-SH2 S03 6 11-30-78 S04 0 06-81 E No Met Towers - Grdg., Ltg. & Details
G374 4 11-30-78 E No Cooling Tower Area - Sections & Details
G376SO1 5 3-15-78 E No Reactor Auxiliary Building Embedded Conduits Below EL(-35) SH1 S02 4 9-30-77 E No Reactor Auxiliary Building Embedded Conduits Below EL(-35) SH2 S03 4 9-30-77 E No Reactor Auxiliary Building Embedded Conduits Below EL(-35) SH3 S04 5 3-15-78 E No Reactor Auxiliary Building Embedded Conduits Below EL(-35) SH4 G377SO1 3 09-77 E No Seismic Cable Supports Piping Penetration Area
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 45 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS
DRAWING & REVISION NUMBER REVISION PREPARED PROPRIETARY TITLE EBASCO OTHERS DATE BY INFO.
203 R3 B-9270-413-101 3 7/13/77 CE No MCBD Temp. Channel 204 3 B-9270-413-102 3 7/13/77 CE No MCBD Temp. Channel 207 3 B-9270-413-105 3 5/30/73 CE No MCBD Press. Pressure Channel 164 4 B-9270-413-107 3 CE No MCBD Press. Pressure Channel 166 3 B-9270-413-109 SH1 3 7/13/77 CE No MCBD Press. & Level Channel 170 4 B-9270-413-201 4 7/13/77 CE No MCBD CVCS Temp. Channel 174 3 B-9270-413-205 3 7/13/77 CE No MCHD CVCS Level Channel 175 2 B-9270-413-207 1 CE No MCBD CVCS Press. Channel 177 4 B-9270-413-209 3 CE No MCBD CVCS Press. Channel 179 3 B-9270-413-301 3 7/13/77 CE No MCBD CVCS SIS Press. & Level Chan- 180 4 B-9270-413-302 3 CE No MCBD CVCS SIS Press. & Level Chan- 181 3 B-9270-413-303 3 7/13/77 CE No MCBD CVCS SIS Press. & Level Chan- 182 4 B-9270-413-304 3 7/13/77 CE No MCBD CVCS SIS Press. & Level Chan- X B-9270-413-304 4 CE No MCBD SIS Flow & Temp. Channel 184 4 B-9270-413-306 4 7/13/77 CE No MCBD SIS Press. & Temp. Chan. 200 4 B-9270-413-401 3 7/13/77 CE No MCBD Stm. Gen. Press & Level Chan. 201 4 B-9270-413-402 3 7/13/77 CE No MCBD Stm. Gen. Press & Level Chan. 718 2 B-9270-413-112 SH1 3 7/13/77 CE No MCBD Rcp Measurement Channel 719 2 SH2 1 7/13/77
720 3 B-9270-413-113 2 4/17/76 721 3 B-9270-413-114 3 7/13/77 CE No MCBD RCP Measurement Channel 751 2 B-9270-413-116 2 7/13/77 CE No MCBD Board Mounted Nuclear Instrumentation SH 1 752 1 B-9270-413-116 1 4/17/76 CE No MCBD Board Mounted Nuclear Instrumentation SH 2 G(DRN 01 758, R11 A) 185 4 B-9270-413-308 3 CE No MCBD RWP Level Channel 0(DRN 01 758, R11 A) 1635 1 D-ICE-411-581 2 5/20/75 CE No Excore Instr- Cabling Diag. 7460 0 E-9270-411-501 1 6/30/77 CE No PPS Functional Diagram SH.1 648 3 E-ICE411-501 1 4/6/76 CE No PPS Functional Diagram SH.2 649 3 E-ICE-411-501 6 7/21/75 CE No PPS Functional Diagram SH.3 8643 0 E-9720-411-503 1 10/7/77 CE No PPS Interface Logic Diagram 663 4 E-ICE-411-520 6 3/29/77 CE No PPS Cab. Assy. & Frt Panel 1/0 SH1 664 4 E-ICE-411-520 4 3/29/77 CE No PPS Cab. Assy. & Frt Panel 1/0 SH2 665 4 E-ICE-411-520 4 3/29/77 CE No PPS Cab. Assy. & Frt Panel 1/0 SH3 7464 0 E-9270-411-521 1 6/30/77 CE No PPS Rmt Control Module 1/0 X J-ICE-411-522 E1 CE No PPS Status - Frt Panel 1/0
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 46 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS
DRAWING & REVISION NUMBER REVISION PREPARED PROPRIETARY TITLE EBASCO OTHERS DATE BY INFO.
667 4 E-ICE-411-531 6 4/78 CE No PPS 2/4 Logic Matrix Sch SH1 668 4 E-ICE-411-531 6 4/78 CE No PPS 2/4 Logic Matrix Sch SH2 669 4 E-ICE-411-531 6 4/78 CE No PPS 2/4 Logic Matrix Sch SH3 670 4 E-ICE-411-531 6 4/78 CE No PPS 2/4 Logic Matrix Sch SH4 671 4 E-ICE-411-531 6 4/78 CE No PPS 2/4 Logic Matrix Sch SH5 672 4 E-ICE-411-531 6 4/78 CE No PPS 2/4 Logic Matrix Sch SH6 7461 0 E-9270-411-532 1 6/30/77 CE No PPS Testing Sys Sch 656 6 E-ICE-411-534 2 4/78 CE No PPS Trip Path Sch 659 5 E-ICE-411-550 4 4/78 CE No PPS Calib & Test Panel Sch 7462 0 E-9270-411-560 1 6/30/77 CE No PPS Bypass & Block Sch 652 3 E-ICE-411-562 4 8/6/75 CE No PPS Misc Sch 7459 0 E-9270-411-570 1 6/30/77 CE No PPS Annum Input Sch Sh 1 646 3 E-9270-411-570 4 7/22/75 CE No PPS Annun Input Sch Sh 2 955 4 E-9270-413-130 E2 CE No Reactor Trip Swgear Arrgmt 1870 1 E-ICE-411-501 3 8/1/75 CE No PPS Functional Diagram Sh 4 1871 2 D-ICE-411-504 3 12/9/76 CE No ESFAS 1875 2 E-ICE-411-537 4 CE No PPS Variable Setpoint Sch 734 2 E-ICE-411-561 3 7/29/74 CE No PPS Aux Logic Sch 1872 1 0-ICE-411-533 2 7/9/75 CE No PPS CPC Test Panel Sch 1873 1 0-ICE-411-533 1 7/9/75 CE No PPS CPC Test Panel Sch 2557 1 B-9270-416-002 1 12/19/76 CE No Swch. Development Sh 1-4 3075 1 B-9270-416-003 1 12/13/76 CE No Power Dist to Instr Cab 1972 1 B-9270-416-007 1 12/13/76 CE No Wide Range Press Press Channel 1977 1 B-9270-416-008 1 12/13/76 CE No Steam Gen. Press Chan 1976 2 B-9270-416-009 SH1 2 CE No Steam Gen. Level Chan SH2 1 SH3 1 SH4 1 1967 2 B-9270-416-010 2 CE No 1978 2 B-9270-416-011 3 7/29/77 CE No RWT Level Chan 1968 1 B-9270-416-012 1 12/13/76 CE No RCP Protective Speed Chan 1969 2 B-9270-416-013 2 CE No Press Press Channel 1970 1 B-9270-416-014 1 12/13/76 CE No Shutdown Heat Exch. 1/0 Temp Chan 8537 0 E-9720-414-075 1 9/23/77 CE No S/D Cooling Temp Channel 653 2 E-9270-411-580 2 CE No CPS Sys Ext Cabling Diagram 654 6 E-9270-411-580 2 CE No PPS Cont Rm Cabling Diagram 1871 2 D-ICE-411-504 3 CE No PPS Cont Rm Cabling Diagram 7553 1 F12004/E30/3ASA 1 CE No ESFAS 6179 0 E-9270-413-021 0 3/31/77 CE No General Layout for E-30 (3A-SA) Steam Bypass Control System
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) G(DRN 02 85, R11 A; 04 1444, R13 B) Start of historical information. 0(DRN 04 1444, R13 B)
WSES-FSAR-UNIT 3 0(DRN 02 85, R11 A) TABLE 1.7-1 (Sheet 47 of 47) Revision 13-B (01/05)
ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS
DRAWING & REVISION NUMBER REVISION PREPARED PROPRIETARY TITLE EBASCO OTHERS DATE BY INFO.
5061 103-524312 Open SEL Block Diagram 5059 142-100067 Open SEL System Layout - Core Protection System (5 Shts.) Interconnection Diagram - Core Protection 5062 144-100555 Open SEL Calculator System (2 Shts.) Cable Assembly - External Operators 5063 144-100556 Open SEL Module Signal (5 Shts.) 5064 149-100316 Open SEL Cable Assembly - External CRT Signal 5066 149-100317 Open SEL CPC "A" Analog Inputs (19 Shts,) CPC "B" Analog Inputs (20 ShLs.) 5067 149-100318 Open SEL CPC "C" Analog Inputs 5068 149-100319 Open SEL CPC "D" Analog Inputs (19 Shts.) 5069 149-100320 Open SEL CEAC "1" Analog Inputs (16 Shts.) 5070 149-100321 Open SEL CEAC "2" Analog Inputs
G(DRN 04 1444, R13 B) End of historical information. 0(DRN 04 1444, R13 B) WSES-FSAR-UNIT-3
1.8 COMPARISON OF WATERFORD 3 DESIGN WITH NRC REGULATORY GUIDES
1.8.1 INTRODUCTION
The Regulatory Guides applicable to Waterford 3 are those referenced in the PSAR. The PSAR and Amendments referenced the Regulatory Guides 1.1 through 1.38 which are relevant to the Waterford design (e.g., PWR).
This section presents a comparison of Waterford 3 plant design with the recommendations presented in the NRC Regulatory Guides 1.1 through 1.96. Regulatory Guide 1.96 Revision 0, May 1975 was issued six months after the Construction Permit date (November 14, 1974) and represents a time when the project was over 50 percent complete in engineering design. A reference to the FSAR sections which the applicable design features are discussed is also provided. Where the design differs from the Regulatory Guide, alternative methods of providing an equivalent level of safety have been utilized; these differences are discussed or reference is made to the appropriate FSAR sections.
In addition, some Regulatory Guides beyond 1.96 have been discussed in the responses to some NRC questions such as 032.3 and 032.6.
1.8.1.1 REGULATORY GUIDE 1.1, NET POSITIVE SUCTION HEAD FOR EMERGENCY CORE COOLING AND CONTAINMENT HEAT REMOVAL PUMPS (Revision 0, November 1970)
Waterford 3 design meets the recommendations of Regulatory Guide 1.1 with the following qualification:
The available NPSH for the safeguard pumps (low pressure and high pressure safety injection pumps and containment spray pumps) is calculated, using a saturated sump model. The containment is assumed to be at the saturation pressure corresponding to the containment sump temperature.
The subject of this Regulatory Guide is discussed in FSAR Subsections 6.2.2.3.2.1 and 6.3.2.2.2.3.
1.8.1.2 REGULATORY GUIDE 1.2, THERMAL SHOCK TO REACTOR PRESSURE VESSELS (Revision 0, November 1970)
Waterford 3 is consistent with the recommendations of Regulatory Guide 1.2. The Heavy Section Steel Technology (HSST) program is discussed in FSAR Subsection 5.3.1.6.
1.8.1.3 REGULATORY GUIDE 1.3, ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A LOSS-OF-COOLANT ACCIDENT FOR BOILING WATER REACTORS (Revision 2, June 1974)
Regulatory Guide 1.3 is not applicable to Waterford 3.
1.8-1
WSES-FSAR-UNIT-3
1.8.1.4 REGULATORY GUIDE 1.4, ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A LOSS-OF-COOLANT ACCIDENT FOR PRESSURIZED WATER REACTORS (Revision 2, June 1974)
(DRN 04-1619, R14) The Loss of Coolant Accident evaluation presented in FSAR Subsection 15.6 utilized the Alternative Source Term dose methodology in accordance with Regulatory Guide 1.183 requirements. As such this Regulatory Guide is no longer applicable to Waterford 3. (DRN 04-1619, R14)
1.8.1.5 REGULATORY GUIDE 1.5, ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A STEAM LINE BREAK ACCIDENT FOR BOILING WATER REACTORS (Revision 0, March 1971)
Regulatory Guide 1.5 is not applicable to Waterford 3.
1.8.1.6 REGULATORY GUIDE 1.6 INDEPENDENCE BETWEEN REDUNDANT STANDBY (ONSITE) POWER SOURCES AND BETWEEN THEIR DISTRIBUTION SYSTEMS (Revision 0, March 1971) (DRN 01-758, R11-A) Waterford 3 design meets the recommendations of Regulatory Guide 1.6. The subject of this Regulatory Guide is discussed in FSAR Subsection 8.3.1.2.3. (DRN 01-758, R11-A) 1.8.1.7 REGULATORY GUIDE 1.7, CONTROL OF COMBUSTIBLE GAS CONCENTRATIONS IN CONTAINMENT FOLLOWING A LOSS-OF-COOLANT ACCIDENT (Revision 0, March 1971)
Waterford 3 design meets the recommendations of Regulatory Guide 1.7. The subject of this Regulatory Guide is discussed in FSAR Subsection 6.2.5.
1.8.1.8 REGULATORY GUIDE 1.8, PERSONNEL SELECTION AND TRAINING (Revision 1, September 1975)
Waterford 3 personnel selection and training are consistent with the recommendations of Regulatory Guide 1.8. The subject of this Regulatory Guide is discussed in FSAR Chapter 13 and QA Program Manual. (LBDCR 14-010, R308) 1.8.1.9 REGULATORY GUIDE 1.9, SELECTION OF DIESEL GENERATOR SET CAPACITY FOR STANDBY POWER SUPPLIES (Revision 0, March 1971 and Revision 4, March 2007)
The subject of this Regulatory Guide is discussed in FSAR Subsection 8.3.1.2.4. IEEE Standard 387 is discussed in FSAR Subsection 8.3.1.2.20. Waterford 3 was originally licensed to Regulatory Guide 1.9 Revision 0. All historic requirements such as the original design criteria, factory production testing, initial type tests, preoperational testing have been completed and approved by the NRC. The historic requirements remain approved to Regulatory Guide 1.9 Revision 0. New or ongoing activities such as surveillance testing, periodic testing, and modifications meet Regulatory Guide 1.9 Revision 4 requirements with the following exceptions and clarifications.
Regulatory Guide 1.9 Revision 4 Clause 1.8 states a trip should be implemented with two or more measurements for each trip parameter with coincident logic provisions for trip actuations. NUREG-0787 Section 8.3.1 provides the NRC review and approval of the current Waterford 3 configuration. For design basis accident conditions, all protective trips except diesel overspeed and generator differential trips are bypassed. (LBDCR 14-010, R308)
1.8-2 Revision 308 11/14
WSES-FSAR-UNIT-3
(LBDCR 14-010, R308; LBDCR 18-018; R311)
Regulatory Guide 1.9 Revision 4 Clause 2.2 states that jumpers and other nonstandard configurations or arrangements should not be used after initial equipment startup testing. Jumpers (nonstandard configuration) are utilized for surveillance tests and are proceduralized to preclude errors. The diesel is not operable during the nonstandard configurations.
Regulatory Guide 1.9 Revision 4 Table 1 shows that protective trip bypass tests are performed at refueling intervals. Regulatory Guide 1.9 Clause 2.2.11 includes the diesel overspeed and generator differential trips in this testing which are performed as part of the maintenance program. The diesel overspeed trip and generator differential trip tests are performed on the diesel maintenance schedule.
Regulatory Guide 1.9 Revision 4 section 2 provides specific surveillance frequencies. Surveillance frequencies contained in the Waterford 3 Technical Specifications and Surveillance Frequency Control Program supersedes the frequencies contained in Regulatory Guide 1.9 Revision 4.
(LBDCR 14-010, R308; LBDCR 18-018, R311)
1.8.1.10 REGULATORY GUIDE 1.10, MECHANICAL (CADWELD) SPLICES IN REINFORCING BARS OF CATEGORY I CONCRETE STRUCTURES (Revision 1, January 1973)
Waterford 3 design meets the recommendations of Regulatory Guide 1.10. The subject of this regulatory guide is discussed in FSAR Section 3.8.
1.8.1.11 REGULATORY GUIDE 1.11, INSTRUMENT LINES PENETRATING PRIMARY REACTOR CONTAINMENT (Revision 0, March 1971)
Waterford 3 design meets the recommendations of Regulatory Guide 1.11. The subject of this Regulatory Guide is discussed in FSAR Subsections 6.2.4.1.3 and 7.1.2.7.
1.8.1.12 REGULATORY GUIDE 1.12, INSTRUMENTATION FOR EARTHQUAKES (Revision 1, April 1974)
Waterford 3 design meets the recommendations of Regulatory Guide 1.12 with the qualifications indicated in FSAR Subsection 3.7.4.1. (DRN 01-758, R11-A) 1.8.1.13 REGULATORY GUIDE 1.13, SPENT FUEL STORAGE FACILITY DESIGN BASES (Revision 0, March 1971) (DRN 01-758, R11-A) Waterford 3 design meets the recommendations of Regulatory Guide 1.13 with the qualification indicated in FSAR Subsection 9.1.4.3. The subject of this Regulatory Guide is discussed in FSAR Subsections 9.1.1, 9.1.2, 9.1.3 and 9.1.4. In FSAR Subsection 15.7.3.4, Revision 1 of Regulatory Guide 1.13 (12/75) is compared with analysis assumptions.
1.8.1.14 REGULATORY GUIDE 1.14 REACTOR COOLANT PUMP FLYWHEEL INTEGRITY (Revision 0, October 1971)
Waterford 3 design meets the recommendations of Regulatory Guide 1.14 for flywheel design and fabrication. The subject of this Regulatory Guide is discussed in FSAR Subsection 5.4.1.4. Technical Specifications discuss flywheel inspection to the requirements of Revision 1 of Regulatory Guide 1.14 (August, 1975).
1.8.1.15 REGULATORY GUIDE 1.15, TESTING OF REINFORCING BARS FOR CATEGORY I CONCRETE STRUCTURES (Revision 1, December 1972)
Waterford 3 meets the recommendations of Regulatory Guide 1.15. The subject of this Regulatory Guide is discussed in FSAR Section 3.8.
1.8.1.16 REGULATORY GUIDE 1.16, REPORTING OF OPERATING INFORMATION APPENDIX A TECHNICAL SPECIFICATIONS (Revision 4, August 1975) 1.8-3 Revision 311 (9/19)
WSES-FSAR-UNIT-3
Waterford 3 meets the recommendations of Regulatory Guide 1.16 to the extent not superseded by 10 CFR 50.73. The subject of this Regulatory Guide is discussed in FSAR Subsection 14.2.7.2 and Technical Specifications.
1.8.1.17 REGULATORY GUIDE 1.17, PROTECTION OF NUCLEAR POWER PLANTS AGAINST INDUSTRIAL SABOTAGE (Revision 0, June 1973)
Waterford 3 meets the recommendations of Regulatory Guide 1.17. The subject of this Regulatory Guide is discussed in the Security Plan.
1.8.1.18 REGULATORY GUIDE 1.18, STRUCTURAL ACCEPTANCE TEST FOR CONCRETE PRIMARY REACTOR CONTAINMENTS (Revision 1, December 1972)
The subject of this Regulatory Guide is not applicable to the Waterford 3 design.
1.8.1.19 REGULATORY GUIDE 1.19 NONDESTRUCTIVE EXAMINATION OF PRIMARY CONTAINMENT LINER WELDS (Revision 1, August 1972)
The subject of this Regulatory Guide is not applicable to The Waterford SES Unit 3 design.
1.8.1.20 REGULATORY GUIDE 1.20, COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM FOR REACTOR INTERNAL DURING PREOPERATIONAL AND INITIAL STARTUP TESTING (Revision 2, May 1976)
Waterford 3 meets the recommendations of Regulatory Guide 1.20. The subject of this Regulatory Guide is discussed in FSAR Subsection 14.2.7.
1.8.1.21 REGULATORY GUIDE 1.21, MEASURING, EVALUATING, AND REPORTING RADIOACTIVITY IN SOLID WASTES AND RELEASES OF RADIOACTIVE MATERIALS IN LIQUID AND GASEOUS EFFLUENTS FROM LIGHT-WATER- COOLED NUCLEAR POWER PLANTS (Revision 1, June 1974)
Waterford 3 meets the recommendations of Regulatory Guide 1.21 as specified in Chapters 11 and 12.
1.8.1.22 REGULATORY GUIDE 1.22, PERIODIC TESTING OF PROTECTION SYSTEM ACTUATION FUNCTIONS (Revision 0, February 1972)
Waterford 3 meets the recommendations of Regulatory Guide 1.22. The subject of this Regulatory Guide is discussed in FSAR Sections 7.2 and 7.3. (EC-1837, R301) 1.8.1.23 REGULATORY GUIDE 1.23, METEOROLOGICAL PROGRAMS IN SUPPORT OF NUCLEAR POWER PLANTS (Proposed Revision 1, September 1980) (EC-1837, R301) Waterford 3 meets the recommendations of Regulatory Guide 1.23. The subject of this Regulatory Guide is discussed in FSAR Subsection 2.3.3.
1.8.1.24 REGULATORY GUIDE 1.24, ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A PRESSURIZED WATER REACTOR RADIOACTIVE GAS STORAGE TANK FAILURE (Revision 0, March 1972) (DRN 04-1619, R14) The Pressurized Water Reactor Radioactive Gas Storage Tank Failure is no longer required to be evaluated by the Standard Review Plan, therefore this event has been deleted from the FSAR. As such this Regulatory Guide is no longer applicable to Waterford 3. (DRN 04-1619, R14)
1.8-4 Revision 301 09/07
WSES-FSAR-UNIT-3
1.8.1.25 REGULATORY GUIDE 1.25, ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCE OF A FUEL HANDLING ACCIDENT IN THE FUEL HANDLING AND STORAGE FACILITY FOR BOILING AND PRESSURIZED WATER REACTORS (Revision 0, March 1972)
(DRN 04-1619, R14) The Fuel Handling Accident evaluation presented in FSAR Subsection 15.7 utilized the Alternative Source Term dose methodology in accordance with Regulatory Guide 1.183 requirements. Thus, Waterford 3 meets the recommendations of Regulatory Guide 1.25 except where superseded by Regulatory Guide 1.183. (DRN 04-1619, R14)
1.8.1.26 REGULATORY GUIDE 1.26, QUALITY GROUP CLASSIFICATIONS AND STANDARDS (Revision 2, June 1975) Waterford 3 meets the recommendations of Regulatory Guide 1.26 with the qualification indicated in FSAR Subsection 3.2.2.
1.8.1.27 REGULATORY GUIDE 1.27, ULTIMATE HEAT SINK FOR NUCLEAR POWER PLANTS (Revision 2, January 1976)
Regulatory Guide 1.27 positions do not exactly apply to the wet-dry cooling tower combination for the Waterford 3 project. However, all concerns raised in the Regulatory Guide are satisfied. A description of the Ultimate Heat Sink is presented in FSAR Subsection 9.2.5.
1.8.1.28 REGULATORY GUIDE 1.28, QUALITY ASSURANCE PROGRAM REQUIREMENTS (DESIGN AND CONSTRUCTION) (Revision 0, June 1972)
Waterford 3 meets the recommendations of Regulatory Guide 1.28.
1.8.1.29 REGULATORY GUIDE 1.29, SEISMIC DESIGN CLASSIFICATION (Revision 1, August 1973)
Waterford 3 design meets the recommendations of Regulatory Guide 1.29 with the qualification indicated in FSAR Subsections 3.2.1 and 8.3.1.2.6.
1.8.1.30 REGULATORY GUIDE 1.30, QUALITY ASSURANCE REQUIREMENTS FOR THE INSTALLATION, INSPECTION, AND TESTING OF INSTRUMENTATION AND ELECTRIC EQUIPMENT (Revision 0, August 1972)
Waterford 3 meets the recommendations of Regulatory Guide 1.30. The subject of this Regulatory Guide is discussed in the QA Program Manual, FSAR Section 7.3 and Subsection 14.2.7.
1.8.1.31 REGULATORY GUIDE 1.31, CONTROL OF STAINLESS STEEL WELDING (Revision 0, August 1972)
The Waterford SES Unit 3 design is consistent with the recommendation of the Interim Position (BTP MTEB 5-1) on Regulatory Guide 1.31 with the qualification indicated in FSAR Subsections 5.2.3.4.2.1 and 6.1.1.1.1.
(EC-2800, R307) The replacement CEDM design is consistent with Regulatory Guide 1.31, Control of Stainless Steel Welding (Revision 3, April 1978). Revision 2 incorporated BTP MTEB 5.1. Revision 3 supersedes prior revisions and BTP MTEB 5-1. (EC-2800, R307)
1.8-5 Revision 307 (07/13)
WSES-FSAR-UNIT-3
1.8.1.32 REGULATORY GUIDE 1.32, CRITERIA FOR SAFETY-RELATED ELECTRIC POWER SYSTEMS FOR NUCLEAR POWER PLANTS (Revision 0, August 1972)
Waterford 3 design meets the recommendations of Regulatory Guide 1.32. The design also complies with positions of Revision 1, dated March 1976, with the qualification indicated in FSAR Subsection 8.3.1.2.8.
1.8.1.33 REGULATORY GUIDE 1.33, QUALITY ASSURANCE PROGRAM REQUIREMENTS (OPERATION) (Revision 2, February 1978)
Waterford 3 meets the recommendations of Regulatory Guide 1.33 with certain exceptions and clarifications as described in the QA Program Manual. The subject of this Regulatory Guide is discussed in FSAR Section 13.5 and the QA Program Manual.
1.8.1.34 REGULATORY GUIDE 1.34, CONTROL OF ELECTROSLAG WELD PROPERTIES (Revision 0, December 1972)
This Regulatory Guide is not applicable to Waterford 3 design.
1.8.1.35 REGULATORY GUIDE 1.35, INSERVICE INSPECTION OF UNGROUTED TENDONS IN PRESTRESSED CONCRETE CONTAINMENT STRUCTURES (Revision 2, January 1976)
This Regulatory Guide is not applicable to Waterford 3 design.
1.8.1.36 REGULATORY GUIDE 1.36, NONMETALLIC THERMAL INSULATION FOR AUSTENITIC STAINLESS STEEL (Revision 0, February 1973)
Waterford 3 design meets the recommendations of Regulatory Guide 1.36. The subject of this Regulatory Guide is discussed in FSAR Subsection 6.1.1.1.4.
1.8.1.37 REGULATORY GUIDE 1.37, QUALITY ASSURANCE REQUIREMENTS FOR CLEANING OF FLUID SYSTEMS AND ASSOCIATED COMPONENTS OF WATER-COOLED NUCLEAR POWER PLANTS (Revision 0, March 1973)
Waterford 3 meets the recommendations of Regulatory Guide 1.37. The subject of this Regulatory Guide is discussed in the QA Program Manual and Subsections 4.5.2.5, 5.4.2, 6.1.1.1.3 and 10.3.6.
1.8.1.38 REGULATORY GUIDE 1.38, QUALITY ASSURANCE REQUIREMENTS FOR PACKAGING, SHIPPING, RECEIVING, STORAGE, AND HANDLING OF ITEMS FOR WATER-COOLED NUCLEAR POWER PLANTS (March, 1973 and Revision 2, May 1977) Waterford 3 meets the recommendations of Regulatory Guide 1.38. The subject of this Regulatory Guide is discussed in the QA Program Manual.
1.8-6 Revision 9 (12/97)
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1.8.1.39 REGULATORY GUIDE 1.39, HOUSEKEEPING REQUIREMENTS FOR WATER-COOLED NUCLEAR POWER PLANTS (Revision 2, September 1977)
-012, R310) Waterford 3 meets the recommendations of Regulatory Guide 1.39. The subject of this Regulatory Guide is discussed in the QA Program Manual and FSAR Subsections 3.8.3 and 14.2.7. CR 16-012, R310)
1.8.1.40 REGULATORY GUIDE 1.40, QUALIFICATION TESTS OF CONTINUOUS- DUTY MOTORS INSTALLED INSIDE THE CONTAINMENT OF WATER COOLED NUCLEAR POWER PLANTS (Revision 0, March 1973) (EC-40281, R307) Waterford 3 meets the recommendations of Regulatory Guide 1.40. Table 6.2-21 indicates motors were qualified to IEEE 334-1974/1994 (EC-40281, R307)
1.8.1.41 REGULATORY GUIDE 1.41, PREOPERATIONAL TESTING OF REDUNDANT ONSITE ELECTRIC POWER SYSTEMS TO VERIFY PROPER LOAD GROUP ASSIGNMENTS (Revision 0, March 1973)
Waterford 3 meets the recommendations of Regulatory Guide 1.41. The subject of this Regulatory Guide is discussed in FSAR Subsections 8.3.1.2.10 and 14.2.7.
1.8.1.42 REGULATORY GUIDE 1.42, INTERIM LICENSING POLICY ON AS LOW AS PRACTICABLE FOR GASEOUS RADIOIODINE RELEASES FROM LIGHT-WATER-COOLED NUCLEAR POWER REACTORS (Revision 1, March 1974)
This Regulatory Guide is considered to be functionally obsolete due to the adoption of the revised Appendix I to 10CFR50. The detailed assumptions for releases, and system supplied, to meet the guidelines of Appendix I to 10CFR50 are given in FSAR Subsections 11.2.3 and 11.3.3.
1.8.1.43 REGULATORY GUIDE 1.43, CONTROL OF STAINLESS STEEL WELD CLADDING OF LOW-ALLOY STEEL COMPONENTS (Revision 0, May 1973)
Waterford 3 meets the recommendations of Regulatory Guide 1.43. The subject of this Regulatory Guide is discussed in FSAR Subsection 5.2.3.3.2.1.
1.8.1.44 REGULATORY GUIDE 1.44, CONTROL OF THE USE OF SENSITIZED STAINLESS STEEL (Revision 0, May 1973)
Waterford 3 design meets the recommendations of Regulatory Guide 1.44. The subject of this Regulatory Guide is discussed in FSAR Subsection 6.1.1.1 and with qualifications as described in Subsection 5.2.3.4.1.1.1.
1.8.1.45 REGULATORY GUIDE 1.45, REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE DETECTION SYSTEMS (Revision 0, May 1973)
(EC-5000082437, R301; EC-19087, R305) Waterford 3 meets the recommendations of Regulatory Guide 1.45 with the following exception and clarification: Only two of the four Leakage Detection methods will meet the sensitivity requirements of Regulatory Guide position C.5. The subject of this Regulatory Guide is discussed in FSAR Subsection 5.2.5.
Note: Revision 1 of RG 1.45 was used to determine the acceptability of the Waterford 3 pressurizer surge line leak-before-break (LBB) analysis as reported in Sections 3.6.3 and 5.2.5. (EC-5000082437, R301; EC-19087, R305)
1.8-7 Revision 310 (12/17)
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1.8.1.46 REGULATORY GUIDE 1.46, PROTECTION AGAINST PIPE WHIP INSIDE CONTAINMENT (May 1973)
(DRN 03-2054, R14) Waterford 3 complies with the Regulatory Guide recommendations for ASME Section III Code Class I pipe for the primary coolant loop, with the following clarification: (EC-19087, R305) Waterford 3 complies with the modified GDC 4 (1987) allowance to credit leak-before- break (LBB) technology to exclude breaks in the primary coolant loop and the pressurizer surge line from consideration of mechanical (dynamic) effects. The next limiting pipe breaks are the remaining branch line pipe breaks (BLPBs), whose effects are evaluated for power uprate to 3716 MWt. A discussion of LBB is contained in Section 3.6.3. (DRN 03-2054, R14; EC-19087, R305)
Waterford 3 complies with the Regulatory Guide recommendations for ASME Section III Code Class I pipe with the following exceptions and clarifications. Piping systems other than primary coolant loop:
Regulatory Position C.1.a - The intersection of two pipes of similar diameters is considered as an intermediate stress point per AEC's letter of July 20, 1973 from Mr. R. Maccary of AEC to Mr. H. Oslick of Ebasco.
Regulatory Position C.1.b - The consideration of pipe breaks during testing is excluded due to the relatively small time period that such testing conditions exist. The exclusion is consistent with the proposed Regulatory Guide for pipe rupture outside the containment, "Protection Against Postulated Piping Failures in Fluid Systems Outside Containment," dated March 15, 1974.
For ASME Section III, Code Class 2 & 3, Waterford 3 has utilized the recommendations of NRC Branch Technical Positions APCSB 3-1, "Protection Against Postulated Piping Failures in Fluid Systems Outside Containment," and MEB 3-1, "Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping," for all piping inside the containment. The subject of this Regulatory Guide is discussed in FSAR Section 3.6. (DRN 06-802, R15) Regulatory Guide 1.46 was withdrawn on March 1, 1985 and superseded with the July 1981 revision of SRP 3.6.2, “Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of piping.” The SRP contained Branch Technical Position (BTP) MEB 3-1 Revision 1, “Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment.” The MEB was revised in Generic Letter (GL) 87-11 (BTP MEB 3-1 Rev. 2), “Relaxation in Arbitrary Intermediate Pipe Rupture Requirements,” to delete the requirements for arbitrary intermediate pipe ruptures. Therefore, Waterford 3 also complies with BTP MEB 3-1 Rev. 2 contained in GL 87-11 for the deletion of arbitrary intermediate pipe ruptures. (DRN 06-802, R15)
(DRN 99-2321, R11) 1.8.1.47 REGULATORY GUIDE 1.47. BYPASSED AND INOPERABLE STATUS INDICATION FOR NULCEAR POWER PLANT SAFETY SYSTEMS. (Revision 0, May 1973)
The subject of this Regulatory Guide is discussed in FSAR Section 7.5.1.8. (DRN 99-2321, R11) 1.8.1.48 REGULATORY GUIDE 1.48, DESIGN LIMITS AND LOADING COMBINATIONS FOR SEISMIC CATEGORY I FLUID SYSTEM COMPONENTS (Revision 0, May 1973)
1.8-8 Revision 305 (11/11)
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The Waterford SES Unit 3 complies with Regulatory Guide 1.48 with the following exceptions and clarifications:
Regulatory Position C.8.a - ASME Code Class 2 & 3 components meet the requirements of the loading combination of Regulatory Position C.8.a.(1). The emergency stress limit of Regulatory Position C.8.a.(2) will not be used. Instead, a stress limit of 1.8 Sh will be used, based on ASME Code Case 1606, November 5, 1973 and AEC Regulatory Guide 1.84, June 1974.
Regulatory Position C.8.b - For the faulted loading combination, Class 2 and 3 components meet the guidance of ASME Section III, Code Case 1606 (November 5, 1973) and AEC Regulatory Guide 1.84, June 1974 which stipulates an allowable stress (at temperature) of 2.4 Sh. The subject of this Regulatory Guide is discussed in FSAR Subsection 3.9.3.
1.8.1.49 REGULATORY GUIDE 1.49, POWER LEVELS OF NUCLEAR POWER PLANTS (Revision 1, December 1973)
Waterford 3 meets the recommendations of Regulatory Guide 1.49.
1.8.1.50 REGULATORY GUIDE 1.50, CONTROL OF PREHEAT TEMPERATURE FOR WELDING OF LOW-ALLOY STEEL (Revision 0, May 1973)
Waterford 3 design meets the recommendations of Regulatory Guide 1.50 with the qualifications indicated in Subsection 5.2.3.3.2.1.
1.8.1.51 REGULATORY GUIDE 1.51, INSERVICE INSPECTION OF ASME CODE CLASS 2 AND 3 NUCLEAR POWER PLANT COMPONENTS (Revision 0, May 1973)
Regulatory Guide 1.51 was withdrawn by the NRC.
1.8.1.52 REGULATORY GUIDE 1.52, DESIGN, TESTING, AND MAINTENANCE CRITERIA FOR ATMOSPHERE CLEANUP SYSTEM AIR FILTRATION AND ADSORPTION UNITS OF LIGHT-WATER-COOLED NUCLEAR POWER PLANTS (Revision 0, June 1973)
Waterford 3 meets the recommendations of Regulatory Guide 1.52 with the qualification indicated in FSAR Table 6.5-1. Revision 0 (6-73) was used for design. Revision 2 (3/78) was used for preoperational, startup and maintenance testing as discussed in FSAR Subsection 14.2.7.9. Revision 2 (3/78) was employed during the startup test program as discussed in FSAR Subsection 14.2.7.9.
1.8.1.53 REGULATORY GUIDE 1.53, APPLICATION OF THE SINGLE-FAILURE CRITERION TO NUCLEAR POWER PLANT PROTECTION SYSTEMS (Revision 0, June 1973)
Waterford 3 design meets the recommendations of Regulatory Guide 1.53. The subject of this Regulatory Guide is discussed in Section 7.2
1.8.1.54 REGULATORY GUIDE 1.54, QUALITY ASSURANCE REQUIREMENTS FOR PROTECTIVE COATINGS APPLIED TO WATER-COOLED NUCLEAR POWER PLANTS (Revision 0, June 1973)
1.8-9 Revision 15 (03/07)
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Waterford 3 meets the recommendations of Regulatory Guide 1.54 with the qualifications indicated in FSAR Subsection 6.1.2.
1.8.1.55 REGULATORY GUIDE 1.55, CONCRETE PLACEMENT IN CATEGORY I STRUCTURES (Revision 0, June 1973)
Waterford 3 design meets the recommendations of Regulatory Guide 1.55 with the qualifications indicated in FSAR Subsection 3.8.3.
1.8.1.56 REGULATORY GUIDE 1.56, MAINTENANCE OF WATER PURITY IN BOILING WATER REACTORS (Revision 0, June 1973)
This Regulatory Guide is not applicable to Waterford 3.
1.8.1.57 REGULATORY GUIDE 1.57, DESIGN LIMITS AND LOADING COMBINATIONS FOR METAL PRIMARY REACTOR CONTAINMENT SYSTEM COMPONENTS (Revision 0, June 1973)
For the steel containment vessel, Waterford 3 project complies with Regulatory Guide with the following exceptions or clarifications:
1. The structural design criteria covers initial and final test conditions, normal operating condition, cold shutdown condition and two postulated accident conditions.
2. The structural design criteria does not include the load combination of post- accident flooding plus 1/2 SSE.
3. Fatigue evaluation is not contemplated in the containment vessel design.
The subject of this Regulatory Guide is discussed in FSAR Subsection 3.8.2.
1.8.1.58 REGULATORY GUIDE 1.58, QUALIFICATION OF NUCLEAR POWER PLANT INSPECTION, EXAMINATION, AND TESTING PERSONNEL (Revision 1, Sept 1980)
Waterford 3 meets the recommendations of Regulatory Guide 1.58. The subject of this Regulatory Guide is discussed in the Quality Assurance Program Manual.
1.8.1.59 REGULATORY GUIDE 1.59, DESIGN BASIS FLOODS FOR NUCLEAR POWER PLANTS (Revision 1, April 1976)
Waterford 3 design meets the recommendations of Regulatory Guide 1.59. The subject of this Regulatory Guide is discussed in FSAR Section 2.4.
1.8.1.60 REGULATORY GUIDE 1.60, DESIGN RESPONSE SPECTRA FOR SEISMIC DESIGN OF NUCLEAR POWER PLANTS (Revision 1, December 1973)
The subject of this Regulatory Guide is discussed in FSAR Subsection 3.7.1.
1.8.1.61 REGULATORY GUIDE 1.61, DAMPING VALUES FOR SEISMIC DESIGN OF NUCLEAR POWER PLANTS (Revision 0, October 1973)
The subject of this Regulatory Guide is discussed in FSAR Section 3.7.
1.8-10 Revision 15 (03/07)
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1.8.1.62 REGULATORY GUIDE 1.62, MANUAL INITIATION OF PROTECTIVE ACTIONS (Revision 0, October 1973)
Waterford 3 meets the recommendations of Regulatory Guide 1.62. The subject of this Regulatory Guide is discussed in FSAR Subsections 7.2.1.1.1.11 and 7.2.2.3.2.
1.8.1.63 REGULATORY GUIDE 1.63, ELECTRIC PENETRATION ASSEMBLIES IN CONTAINMENT STRUCTURES FOR WATER-COOLED NUCLEAR POWER PLANTS (Revision 0, October 1973)
Waterford 3 design meets the recommendations of Regulatory Guide 1.63. The subject of this Regulatory Guide is discussed in FSAR Subsections 8.3.1.1.4 and 14.2.7.
1.8.1.64 REGULATORY GUIDE 1.64, QUALITY ASSURANCE REQUIREMENTS FOR THE DESIGN OF NUCLEAR POWER PLANTS (Revision 2, June 1976)
Waterford 3 meets the recommendations of Regulatory Guide 1.64. The subject of this Regulatory Guide is discussed in FSAR Section 17.2.
1.8.1.65 REGULATORY GUIDE 1.65, MATERIALS AND INSPECTIONS FOR REACTOR VESSEL CLOSURE STUDS (Revision 0, October 1973)
The subject of this Regulatory Guide is discussed in FSAR Subsections 5.3.1.7 and 14.2.7.
1.8.1.66 REGULATORY GUIDE 1.66, NONDESTRUCTIVE EXAMINATION OF TUBULAR PRODUCTS (Revision 0, October 1973)
Waterford 3 meets the recommendations of Regulatory Guide 1.66 with the qualifications indicated in FSAR Subsection 5.2.3.3.2.4.
1.8.1.67 REGULATORY GUIDE 1.67, INSTALLATION OF OVERPRESSURE PROTECTION DEVICES (Revision 0, October 1973)
Waterford 3 design meets the recommendations of Regulatory Guide 1.67.
1.8.1.68 REGULATORY GUIDE 1.68, PREOPERATIONAL AND INITIAL STARTUP TEST PROGRAMS FOR WATER-COOLED POWER REACTORS (Revision 2, August 1978)
Waterford 3 meets the recommendations of Regulatory Guide 1.68 with the qualification indicated in FSAR Subsection 14.2.7.13.
1.8.1.69 REGULATORY GUIDE 1.69, CONCRETE RADIATION SHIELDS FOR NUCLEAR POWER PLANTS (Revision 0, December 1973)
Waterford 3 meets the recommendations of Regulatory Guide 1.69 with the qualification indicated in FSAR Subsection 12.3.2.4.
1.8-11
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1.8.1.70 REGULATORY GUIDE 1.70, STANDARD FORMAT AND CONTENT OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS (Revision 2, September 1975)
The recommendations of Regulatory Guide 1.70, Revision 2, were followed in preparing this FSAR.
1.8.1.71 REGULATORY GUIDE 1.71, WELDER QUALIFICATION FOR AREAS OF LIMITED ACCESSIBILITY (Revision 0, December 1973)
The subject of this Regulatory Guide is discussed in FSAR Subsections 5.2.3.3.2.3, 6.1.1.1.3 and 4.5.2.4.5 and with qualifications as discussed in Subsection 10.3.6.
1.8.1.72 REGULATORY GUIDE 1.72, SPRAY POND PLASTIC PIPING (Revision 0, December 1973)
This Regulatory Guide is not applicable to Waterford 3.
1.8.1.73 REGULATORY GUIDE 1.73, QUALIFICATION TESTS OF ELECTRIC VALVE OPERATORS INSTALLED INSIDE THE CONTAINMENT OF NUCLEAR POWER PLANTS (Revision 0, January 1974)
Waterford 3 meets the recommendations of Regulatory Guide 1.73.
1.8.1.74 REGULATORY GUIDE 1.74, QUALITY ASSURANCE TERMS AND DEFINITIONS (Revision 0, February 1974)
Waterford 3 meets the recommendations of Regulatory Guide 1.74. The subject of this Regulatory Guide is discussed in the QA Program Manual.
1.8.1.75 REGULATORY GUIDE 1.75, PHYSICAL INDEPENDENCE OF ELECTRICAL SYSTEMS (Revision 1, January 1975)
BDCR 16-012, R310) Waterford 3 SES Unit 3 design ensures that separation criteria as required by Regulatory Guide 1.75 is met. The subject of this Regulatory Guide is discussed in FSAR Subsection 8.3.1.2.13. -012, R310)
1.8.1.76 REGULATORY GUIDE 1.76, DESIGN BASIS TORNADO FOR NUCLEAR POWER PLANTS (Revision 0, April 1974)
Waterford 3 design basis tornado parameters are presented in FSAR Subsection 3.3.2.
1.8.1.77 REGULATORY GUIDE 1.77, ASSUMPTIONS USED FOR EVALUATING A CONTROL ROD EJECTION ACCIDENT FOR PRESSURIZED WATER REACTORS (Revision 0, May 1974)
(DRN 04-1619, R14) The Control Rod Ejection evaluation presented in FSAR Subsection 15.4 utilized the Alternative Source Term dose methodology in accordance with Regulatory Guide 1.183 requirements. Thus, Waterford 3 meets the recommendations of Regulatory Guide 1.77 except where superseded by Regulatory Guide 1.183. (DRN 04-1619, R14)
1.8-12 Revision 310 (12/17)
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1.8.1.78 REGULATORY GUIDE 1.78, ASSUMPTIONS FOR EVALUATING THE HABITABILITY OF A NUCLEAR POWER PLANT CONTROL ROOM DURING A POSTULATED HAZARDOUS CHEMICAL RELEASE (Revision 0, June 1974)
The subject of this Regulatory Guide is discussed in FSAR Subsections 2.2.3.3 and 6.4.4.2.
1.8.1.79 REGULATORY GUIDE 1.79, PREOPERATIONAL TESTING OF EMERGENCY CORE COOLING SYSTEMS FOR PRESSURIZED WATER REACTORS (Revision 1, September 1975)
Waterford 3 meets the recommendation of Regulatory Guide 1.79 with the qualification indicated in FSAR Subsection 14.2.7.15.
1.8.1.80 REGULATORY GUIDE 1.80, PREOPERATIONAL TESTING OF INSTRUMENT AIR SYSTEMS (Revision 0, June 1974)
Waterford 3 will meet the recommendations of Regulatory Guide 1.80 with the qualifications indicated in Subsection 14.2.7.16.
1.8.1.81 REGULATORY GUIDE 1.81, SHARED EMERGENCY AND SHUTDOWN ELECTRIC SYSTEMS FOR MULTI-UNIT NUCLEAR POWER PLANTS (Revision 1, January 1975)
The subject of this Regulatory Guide is not applicable to Waterford 3.
1.8.1.82 REGULATORY GUIDE 1.82, SUMPS FOR EMERGENCY CORE COOLING AND CONTAINMENT SPRAY SYSTEMS (Revision 0, June 1974)
Waterford 3 design meets the recommendations of Regulatory Guide 1.82 with the qualifications indicated in FSAR Subsection 6.2.2.2.2.1.
1.8.1.83 REGULATORY GUIDE 1.83, IN-SERVICE INSPECTION OF PRESSURIZED WATER REACTOR STEAM GENERATOR TUBES (Revision 1, July 1975) (EC-8458, R307) Regulatory Guide 1.83 (Revision 1) was withdrawn on November 12, 2009. Inservice Inspection of steam generator tubes is performed in accordance with the Technical Specifications. (EC-8458, R307)
1.8.1.84 REGULATORY GUIDE 1.84, CODE CASE ACCEPTABILITY-ASME SECTION III DESIGN AND FABRICATION
Waterford 3 design meets the recommendations of Regulatory Guide 1.84 with the qualifications indicated in FSAR Subsections 5.2.1.2.1 and 3.8.2.
1.8.1.85 REGULATORY GUIDE 1.85, CODE CASE ACCEPTABILITY-ASME SECTION III MATERIALS
Waterford 3 meets the recommendations of Regulatory Guide 1.85 with the qualifications indicated in FSAR Subsections 5.2.1.2.2 and 3.8.2.
1.8-13 Revision 307 (07/13)
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1.8.1.86 REGULATORY GUIDE 1.86, TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS (June 1974)
Waterford 3 project will meet the recommendations of Regulatory Guide 1.86, or other guidance extant at the time of decommissioning.
1.8.1.87 REGULATORY GUIDE 1.87, CONSTRUCTION CRITERIA FOR CLASS I COMPONENTS IN ELEVATED TEMPERATURE REACTORS (SUPPLEMENT TO ASME SECTION III, CODE CASES 1592, 1593, 1594, 1595 and 1596) (Revision 1, June 1975)
The subject of this Regulatory Guide is not applicable to the Waterford 3 design.
1.8.1.88 REGULATORY GUIDE 1.88, COLLECTION, STORAGE AND MAINTENANCE OF NUCLEAR POWER PLANT QUALITY ASSURANCE RECORDS (Revision 2, October 1976)
(LBDCR 16-054, R310) Waterford 3 meets the recommendations of Regulatory Guide 1.88 or ANSI/ASME NQA-1 with the qualifications indicated in FSAR section 13.4 and clarification/exception provided in the Quality Assurance Program Manual. The subject of this Regulatory Guide is discussed in FSAR Subsection 14.2.7.17 and QA Program Manual. (LBDCR 16-054, R310)
1.8.1.89 REGULATORY GUIDE 1.89, QUALIFICATION OF CLASS 1E EQUIPMENT FOR NUCLEAR POWER PLANTS (November 1974)
Waterford 3 meets the recommendations of Regulatory Guide 1.89. The subject of this Regulatory Guide is discussed in FSAR Subsection 8.3.1.2.
1.8.1.90 REGULATORY GUIDE 1.90, INSERVICE INSPECTION OF PRESTRESSED CONCRETE CONTAINMENT STRUCTURES WITH GROUTED TENDONS (November 1974)
The subject of this Regulatory Guide is not applicable to the Waterford 3.
1.8.1.91 REGULATORY GUIDE 1.91, EVALUATION OF EXPLOSIONS POSTULATED TO OCCUR ON TRANSPORTATION ROUTES NEAR NUCLEAR POWER PLANT SITES (January 1975)
Waterford 3 meets the recommendations of Regulatory Guide 1.91.
1.8.1.92 REGULATORY GUIDE 1.92, COMBINATION OF MODES AND SPATIAL COMPONENTS IN SEISMIC RESPONSE ANALYSIS (Revision 0, December 1974)
The mode and spatial component combinations used are discussed in FSAR Subsections 3.7.1, 3.7.2, and 3.7.3.
1.8.1.93 REGULATORY GUIDE 1.93, AVAILABILITY OF ELECTRIC POWER SOURCES (December 1974)
Waterford 3 meets the recommendations of Regulatory Guide 1.93.
1.8-14 Revision 310 (12/17)
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1.8.1.94 REGULATORY GUIDE 1.94, QUALITY ASSURANCE REQUIREMENTS FOR INSTALLATION, INSPECTION, AND TESTING OF STRUCTURAL CONCRETE AND STRUCTURAL STEEL DURING THE CONSTRUCTION PHASE OF NUCLEAR POWER PLANTS (Revision 1, April 1976)
The quality assurance requirements for installation, inspection, and testing of structural concrete and structural steel during construction are discussed in QA Program Manual. (DRN 01-758, R11-A) 1.8.1.95 REGULATORY GUIDE 1.95, PROTECTION OF NUCLEAR POWER PLANT CONTROL ROOM OPERATORS AGAINST AN ACCIDENTAL CHLORINE RELEASE (February 1975) (DRN 01-758, R11-A) Waterford 3 meets the recommendations of Regulatory Guide 1.95. The subject of this Regulatory Guide is discussed in FSAR Subsections 2.2.3.3, 6.4.4.2, and 14.2.7.18.
1.8.1.96 REGULATORY GUIDE 1.96, DESIGN OF MAIN STEAM ISOLATION VALVE LEAKAGE CONTROL SYSTEMS FOR BOILING WATER REACTOR NUCLEAR POWER PLANTS (Revision 1, June 1976)
The subject of this Regulatory Guide is not applicable to Waterford 3. (DRN 01-758, R11-A) 1.8.1.97 REGULATORY GUIDE 1.97, INSTRUMENTATION FOR LIGHT-WATER-COOLED NUCLEAR POWER PLANTS TO ASSESS PLANT AND ENVIRONS CONDITIONS DURING AND FOLLOWING AN ACCIDENT (Revision 3, December 1983)
(DRN 01-758, R11-A) (DRN 01-26, R11-A; 01-758, R11-A; ) A complete report of the compliance of Waterford 3 with the recommendations of Regulatory Guide 1.97 and the schedule of completion of the various instruments is provided to the NRC in letter W3F1-91-0019. This letter was supplemented by letter W3F1-97-0120. Additional information regarding conformance to Regulatory Guide 1.97 can be found in Section 7.5.1.7. (DRN 01-26, R11-A; 01-758, R11-A) (DRN 01-758, R11-A; 01-1281, R12)
→ (LBDCR 19-013, R311) 1.8.1.99 REGULATORY GUIDE 1.99 RADIATION EMBRITTLEMENT OF REACTOR VESSEL MATERIALS (Revision 2, May 1988)
Waterford 3 utilized the methods provided in RG 1.99 for evaluation of reactor vessel neutron embrittlement as described in UFSAR Subsection 18.2.1 ← (LBDCR 19-013, R311)
1.8.1.109 CALCULATION OF ANNUAL DOSE TO MAN FROM ROUTINE RELEASES OF REACTOR EFFLUENTS FOR THE PURPOSE OF EVALUATING COMPLIANCE WITH 10 CFR PART 50, APPENDIX I (March 1976)
Waterford 3 meets the recommendations of Regulatory Guide 1.109 as discussed in FSAR Subsections 11.2.3, 11.3.3, 12.2.2, and 15.7.3.15.1.
1.8.1.111 METHODS FOR ESTIMATING ATMOSPHERIC TRANSPORT AND DISPERSION OF GASEOUS EFFLUENTS IN ROUTINE RELEASES FROM LIGHT-WATER-COOLED REACTORS (March 1976)
Waterford 3 meets the recommendations of Regulatory Guide 1.111 as discussed in FSAR Subsection 2.3.5.1. (DRN 01-1281, R12)
1.8-15 Revision 311 (9/19)
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EC-26965, R305) 1.8.1.133 LOOSE PART DETECTION PROGRAM FOR THE PRIMARY SYSTEM OF LIGHT WATER COOLED REACTOR
Waterford 3 meets the recommendations of Regulatory Guide 1.133 as discussed in UFSAR Subsection 4.4.6.1 and modified by NRC Technical Specification Amendment 104. (EC-26965, R305) 1.8.1.143 REGULATORY GUIDE 1.143, DESIGN GUIDANCE FOR RADIOACTIVE WASTE MANAGEMENT SYSTEMS, STRUCTURES, AND COMPONENTS INSTALLED IN LIGHT-WATER-COOLED NUCLEAR POWER PLANTS (Revision 1, October 1979) (EC-47424, R308) The liquid, gaseous and solid waste management systems conform with Regulatory Guide 1.143 requirements with the following exception. The piping and valve design, construction, inspection and testing requirements meet ANSI B31.1 and/or ANSI B31.3 standards. (EC-47424, R308) (DRN 03-2054, R14) 1.8.1.145 REGULATORY GUIDE 1.145, ATMOSPHERIC DISPERSION MODELS FOR POTENTIAL ACCIDENT CONSEQUENCE ASSESSMENTS AT NUCLEAR POWER PLANTS (REVISION 0, AUGUST 1979)
Waterford 3 meets the recommendations of Regulatory Guide 1.145 as discussed in FSAR Subsection 2.3.4.
→ (LBDCR 19-013, R311) 1.8.1.160 REGULATORY GUIDE 1.160, MONITORING THE EFFECTIVENESS OF MAINTENANCE AT NUCLEAR POWER PLANTS (Revision 2, March 1997)
The Structures Monitoring Program described in UFSAR Subsection 18.1.38 was developed based on the guidance in RG 1.160. ← (LBDCR 19-013, R311)
(DRN 03-2054, R14) (DRN 04-1619, R14) 1.8.1.183 REGULATORY GUIDE 1.183, ALTERNATIVE RADIOLOGICAL SOURCE TERMS FOR EVALUATING DESIGN BASIS ACCIDENTS AT NUCLEAR POWER REACTORS (Revision 0, July 2000)
The design basis accident dose analyses contained in Chapter 15 utilized the Alternative Source Term methodology in accordance with Regulatory Guide 1.183 requirements. Waterford 3 meets or exceeds the recommendations of this Regulatory Guide with the following exceptions and clarifications.
The breathing rates assumed for the control room and off-site doses uses a more accurate value of 3.47E-4 m3/s versus the rounded up 3.5E-4 m3/s specified in the regulatory guidance.
Regulatory Guide 1.183, Section 5.1 of Appendix E states “For facilities with traditional generator specifications (both per generator and total of all generators), the leakage should be apportioned between the affected and unaffected steam generators in such a manner that the calculated dose is maximized.” For the accident sequences presented in Chapter 15, the primary-to-secondary side steam generator tube leakage is specified on a per steam generator basis.
Regulatory Guide 1.183, Appendix A, recommends that a flashing fraction of 10% be applied to the ESF liquid leakage term assumed in the loss of coolant accident dose analysis for the duration of the event. Waterford 3 conservatively uses a value of 10% for the first 24 hours of the event, however, the value is reduced to 2% thereafter. The 2% value is roughly a factor of 10 greater than what would be expected based on the post-LOCA safety injection sump temperature profile. (DRN 04-1619, R14)
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(LBDCR 17-007, R310; LBDCR 19-013, R311) 1.8.1.190 REGULATORY GUIDE 1.190, CALCULATIONAL AND DOSIMETRY METHODS FOR DETERMINING VESSEL NEUTRON FLUENCE (March 2001)
As described in UFSAR Subsection 18.2.1.1 the methods used to calculate the reactor vessel neutron fluence satisfy the criteria set forth in RG 1.190. ←(LBDCR 19-013, R311)
1.8.1.199 REGULATORY GUIDE 1.199, ANCHORING COMPONENTS AND STRUCTURAL SUPPORTS IN CONCRETE (November 2003)
Regulatory Guide 1.199 endorses ACI 349-01 Appendix B. Waterford 3 meets the requirements for implementing ACI 349-01 Appendix B as described in FSAR Subsection 3.8.3.2. (LBDCR 17-007, R310)
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1.9 THREE MILE, ISLAND - 2 (TMI-2) ACTION PLAN REQUIREMENTS FOR APPLICANTS FOR AN OPERATING LICENSE
On October 31, 1980, D. G. Eisenhut, Director, Division of Licensing, Office of Nuclear Reactor Regulation, issued a letter to "All Licensees of Operating Plants and Applicants for Operating Licenses and Holders of Construction Permits" addressing Post TMI - requirements (NUREG-0737). Enclosure 2 to this document identified TMI Action Plan Requirements for Applicants for an Operating License approved for implementation by the Commission at the time of issuance.
In this section, those specific requirements of enclosure 2, as cited above, which affect Waterford 3 are identified and addressed. Additionally, Table 1.9-1 provides a reference to an FSAR section or sections where Waterford 3's method of compliance is described.
1.9.1 SHIFT TECHNICAL ADVISOR (I.A.1.1.)
Position
Each licensee shall provide an on-shift technical advisor to the shift supervisor. The shift technical advisor (STA) may serve more than one unit at a multiunit site if qualified to perform the advisor function for the various units.
The STA shall have a bachelor's degree or equivalent in a scientific or engineering discipline and have received specific training in the response and analysis of the plant for transients and accidents. The STA shall also receive training in plant design and layout, including the capabilities of instrumentation and controls in the control room. The licensee shall assign normal duties to the STAs that pertain to the engineering aspects of assuring safe operations of the plant, including the review and evaluation of operating experience.
Response
Technical Advisors with engineering expertise and special training in plant dynamic response will be available on shift to advise and assist the Shift Supervisor in the event of an accident. Subsection 13.1.2.1.5.1 of the FSAR has been expanded to discuss the details of this commitment.
STA requirements are specified in the Technical Specifications.
1.9.2 SHIFT SUPERVISOR ADMINISTRATIVE DUTIES (I.A.1.2)
Position
Review the administrative duties of the shift supervisor and delegate functions that detract from or are subordinate to the management responsibility for assuring safe operation of the plant to other personnel not on duty in the control room.
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Response
(LBDCR 13-015, R308) A review of the Shift Supervisor's (Shift Manager’s) duties has been conducted to relieve him of those administrative functions that detract from or are subordinate to his management responsibility for assuring the safe operation of the plant. These administrative duties are delegated as appropriate to other operations personnel not on duty in the control room. (LBDCR 13-015, R308)
1.9.3 SHIFT MANNING (I.A.1.3)
Position
This position defines shift manning requirements for normal operation. The letter of July 31, 1980 from D. G. Eisenhut to all power reactor licensees and applicants sets forth the interim criteria for shift staffing (to be effective pending general criteria that will be the subject of future rulemaking). Overtime restrictions were also included in the July 31, 1980 letter.
Response
Shift manning requirements are specified in the Technical Specifications.
1.9.4 IMMEDIATE UPGRADING OF REACTOR OPERATOR AND SENIOR REACTOR OPERATOR TRAINING AND QUALIFICATIONS (I.A.2.1)
Position
Effective December 1, 1980, an applicant for a senior reactor operator (SRO) license will be required to have been a licensed operator for 1 year.
Response
Licensed personnel training qualification requirements are discussed in 2 Subsection 13.2.1.2 of the FSAR.
Licensed operators have been screened and their job positions analyzed using position task analyses to determine training requirements. FSAR Subsection 13.2.1 has been revised to reflect this commitment.
1.9.5 ADMINISTRATION OF TRAINING PROGRAMS (I.A.2.3)
Position
Pending accreditation of training institutions, licensees and applicants for operating licenses will assure that training center and facility instructors who teach systems, integrated responses, transient, and simulator courses demonstrate senior reactor operator (SRO) qualifications and be enrolled in appropriate requalification programs.
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Response (DRN 01-758) The associated Waterford 3 Training Programs are based on a systematic approach to training. The Licensed Operator and Shift Technical Advisor Programs were initially accredited by INPO February 25, 1987. Instructors who routinely teach systems important to plant safety, integrated responses, transient and simulator courses have demonstrated SRO qualifications and are enrolled in appropriate requalification programs. (DRN 01-758) 1.9.6 REVISE SCOPE AND CRITERIA FOR LICENSING EXAMINATIONS (I.A.3.1)
Position
All reactor operator license applicants shall take a written examination with a new category dealing with the principles of heat transfer and fluid mechanics, a time limit of nine hours, and a passing grade of 80 percent overall and 70 percent in each category.
All senior reactor operator license applicants shall take the reactor operator examination, an operating test, and a senior reactor operator written examination with a new category dealing with the theory of fluids and thermodynamics, a time limit of seven hours, and a passing grade of 80 percent overall and 70 percent in each category.
Simulator examinations will be included as part of the licensing examinations.
Response
The scope and criteria for requalification of operating personnel were to reflect the added requirements of: a. Instruction in heat transfer, fluid flow, thermodynamics and mitigation of accidents involving a degraded core. b. Accelerated requalification quiz and exam grade levels. c. Specified reactivity manipulations.
1.9.7 INDEPENDENT SAFETY ENGINEERING GROUP (I.B.1.2)
Position
Each applicant for an operating license shall establish an onsite independent safety engineering group (ISEG) to perform independent reviews of plant operations.
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The principal function of the ISEG is to examine plant operating characteristics, NRC issuances, Licensing Information Service advisories, and other appropriate sources of plant design and operating experience information that may indicate areas for improving plant safety. The ISEG is to perform independent review and audits of plant activities including maintenance, modifications, operational problems, and operational analysis, and aid in the establishment of programmatic requirements for plant activities. Where useful improvements can be achieved, it is expected that this group will develop and present detailed recommendations to corporate management for such things as revised procedures or equipment modifications.
Another function of the ISEG is to maintain surveillance of plant operations and maintenance activities to provide independent verification that these activities are performed correctly and that human errors are reduced as far as practicable. ISEG will then be in a position to advise utility management on the overall quality and safety of operations. ISEG need not perform detailed audits of plant operations and shall not be responsible for sign-off functions such that it becomes involved in the operating organization.
Response
(DRN 00-576, R11; LBDCR 13-015, R308) The Independent Technical Review (ITR) function is no longer a specific designated function for plant operating oversight and reduction of human errors. Rather, the oversight function is performed as part of on-going processes for assessing plant operation at Waterford 3. Those functions include activities conducted by Oversight, Performance Improvement, Regulatory Assurance, and Engineering. Human performance improvement has been integrated into all site functions and is a goal for all departments. The combination of these various activities meets the intent for independent safety review for the commitment to NUREG-0737, Section l.B.1.2, as follows: (LBDCR 13-015, R308)
An operations experience group evaluates and distributes in-house and industry information to appropriate EOI personnel for review. Recommendations resulting from these reviews are implemented to improved reliability and safety. Design engineering support for Waterford 3 is located on site, making the engineers readily available to address potential design basis issues. Plant engineering support for Waterford 3 is located on site and is responsible for optimizing system performance and reliability and for providing technical assistance to the Operations and Maintenance organizations. The corrective action program contains the essential process elements of problem reporting, root cause analysis, and corrective action. The use of assessments provides information on performance trends and improvements for EOI and Waterford 3 management. (DRN 03-657, R12-C) Oversight committees (SRC and OSRC) review plant operations. (DRN 03-657, R12-C) (LBDCR 13-015, R308) Management participation in the Performance Improvement program process (e.g., review of condition reports, grading the significance of condition reports, review of root cause analyses, and determination of which conditions relate to human performance) ensures that the quality and integrity of the program is maintained and that problems are visible to Waterford 3 management. (DRN 00-576, R11; LBDCR 13-015, R308)
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1.9.8 SHORT-TERM ACCIDENT ANALYSIS AND PROCEDURE REVISION (I.C.1)
Position
Analyze small-break LOCAs over a range of break sizes, locations and conditions (including some specified multiple equipment failures) and inadequate core cooling due to both low reactor coolant system inventory and the loss of natural circulation to determine the important phenomena involved and expected instrument indications. Based on these analyses, revise as necessary emergency procedures and training.
Response
A small break LOCA analysis has been conducted on Waterford 3 and is discussed in Sections 6.3 and 15.6. Additionally, LP&L has participated in the CE Owners Group effort to develop Emergency Procedure Guidelines (EPGs) which are based, in part, on generic analyses of small-break LOCAs. The Waterford 3 emergency operating procedures implement the CE Owners Group EPGs.
Cold license candidate’s onsite training includes training in heat transfer, fluid flow and thermodynamics. It also includes comprehensive training in the use of installed plant systems to control or mitigate an accident in which the core is severely damaged. Subsection 13.2.1 of the FSAR discusses this training. The “contingency plan” for cold license candidates calls for the review and discussion of Operating, Emergency and Abnormal Procedures by operational personnel (also in Subsection 13.2.1). This review and discussion when combined with the previously mentioned training should prevent procedural inadequacies prior to implementation of the procedure.
1.9.9 SHIFT RELIEF AND TURNOVER PROCEDURES (I.C.2)
Position
Revise plant procedures for shift relief and turnover to require signed checklists and logs to assure that the operating staff (including auxiliary operators and maintenance personnel) possess adequate knowledge of critical plant parameter status, system status, availability and alignment.
Response (DRN 02-559, R12, LBDCR 15-027, R309) Entergy Operations Procedures, such as EN-OP-115 (and its associated progeny procedures), entitled “Conduct of Operations”, define the responsibilities and methods to be used by the Operations Group to ensure that plant operations are conducted in conformance with applicable legal requirements and regulations and dictates of good operating practices. (DRN 02-559, R12, LBDCR 15-027, R309) (LBDCR 13-015, R308) 1.9.10 SHIFT MANAGER RESPONSIBILITIES (I.C.3) (LBDCR 13-015, R308) Position
Issue a corporate management directive that clearly establishes the command duties of the shift supervisor and emphasizes the primary management responsibility for safe operation of the plant. Revise plant procedures to clearly define the duties, responsibilities and authority of the shift supervisor and the control room operators.
Response (DRN 02-559, R12. LBDCR 15-027, R309) Entergy Operations Procedures, such as EN-OP-115 (and its associated progeny procedures), entitled “Conduct of Operations”, define the responsibilities and methods to be used by the Operations Group to ensure that plant operations are conducted in conformance with applicable legal requirements and regulations and dictates of good operating practices. A corporate management directive concerning command duties of the shift supervisor is issued yearly in accordance with the Technical Specifications. (DRN 02-559, R12, LBDCR 15-027. R309) 1.9-5 Revision 309 (06/16)
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(LBDCR 13-015, R308) and dictates of good operating practices. A corporate management directive concerning command duties of the shift manager is issued yearly in accordance with the Technical Specifications. (LBDCR 13-015, R308) 1.9.11 CONTROL ROOM ACCESS (I.C.4)
Position
Revise plant procedures to limit access to the control room to those individuals responsible for the direct operation of the plant, technical advisors, specified NRC personnel, and to establish a clear line of authority, responsibility, and succession in the control room.
Response (DRN 02-559, R12, LBDCR 15-027. R309) Entergy Operations Procedures, such as EN-OP-115 (and its associated progeny procedures), entitled “Conduct of Operations”, define the responsibilities and methods to be used by the Operations Group to ensure that plant operations are conducted in conformance with applicable legal requirements and regulations and dictates of good operating practices. (DRN 02-559, R12, LBDCR 15-027, R309) 1.9.12 PROCEDURES FOR FEEDBACK OF OPERATING EXPERIENCE TO PLANT STAFF (I.C.5)
Position
Each applicant for an operating license shall prepare procedures to assure that operating information pertinent to plant safety originating both within and outside the utility organization is continually supplied to operators and other personnel and is incorporated into training and retraining programs. These procedures shall:
(1) Clearly identify organizational responsibilities for review of operating experience the feedback of pertinent information to operators and other personnel, and the incorporation of such information into training and retraining programs;
(2) Identify the administrative and technical review steps necessary in translating recommendations by the operating experience assessment group into plant actions (e.g., changes to procedures; operating orders);
(3) Identify the recipients of various categories of information from operating experience (i.e., supervisory personnel, shift technical advisors, operators, maintenance personnel, health physics technicians) or otherwise provide means through which such information can be readily related to the job functions of the recipients;
(4) Provide means to assure that affected personnel become aware of and understand information of sufficient importance that should not wait for emphasis through routine training and retraining programs;
(5) Assure that plant personnel do not routinely receive extraneous and unimportant information on operating experience in such volume that it would obscure priority information or otherwise detract from overall job performance and proficiency;
(6) Provide suitable checks to assure that conflicting or contradictory information is not conveyed to operators and other personnel until resolution is reached; and,
(7) Provide periodic internal audit to assure that the feedback program functions effectively at all levels.
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Response
Administrative Procedures ensure that operating experience from within and outside the LP&L organization is provided to operators and other operating personnel and is incorporated in training programs in accordance with NRC instructions. Waterford 3 Nuclear Training Procedure NTP-102, entitled "Licensed Reactor Operator Requalification" provides the program description and provides instructions for the conduct and documentation of operator retraining. Prior to the cold license examination this program shall be utilized to maintain the proficiency and basic levels of knowledge of the licensed personnel.
1.9.13 GUIDANCE ON PROCEDURES FOR VERIFYING CORRECT PERFORMANCE OF OPERATING ACTIVITIES (I.C.6)
Position
It is required that licensees' procedures be reviewed and revised, as necessary, to assure that an effective system of verifying the correct performance of operating activities is provided as a means of reducing human errors and improving the quality of normal operations. This will reduce the frequency of occurrence of situations that could result in or contribute to accidents. Such a verification system may include automatic system status monitoring, human verification of operations and maintenance activities independent of the people performing the activity or both.
Implementation of automatic status monitoring if required will reduce the extent of human verification of operations and maintenance activities but will not eliminate the need for such verification in all instances. The procedures adopted by the licensees may consist of two phases--one before and one after installation of automatic status monitoring equipment, if required.
Response (DRN 02-559, R12, LBDCR 15-027, R309) Entergy Operations Procedures, such as EN-OP-115 (and its associated progeny procedures), entitled “Conduct of Operations”, define the responsibilities and methods to be used by the Operations Group to ensure that plant operations are conducted in conformance with applicable legal requirements and regulations and dictates of good operating practices. (DRN 02-559, R12, LBDCR 15-027, R309) 1.9.14 NSSS VENDOR REVIEW OF PROCEDURES (I.C.7)
Position
Obtain nuclear steam supply system (NSSS) vendor review of low-power testing procedures to further verify their adequacy.
Response
Emergency Operating Procedures were developed using the Combustion Engineering Emergency Procedure Guidelines (EPGs) as their basis. Use of the EPGs satisfies the I.C.7 requirement for vendor review.
A detailed review and independent analysis of both low-power and power ascension test procedures was conducted to verify the adequacy of these procedures. Additionally the NSSS vendor assisted in analyzing and interpreting the testing program results.
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1.9.15 CONTROL ROOM DESIGN REVIEWS (I.D.1)
Position
All licensees and applicants for operating licenses will be required to conduct a detailed control room design review to identify and correct design deficiencies. This detailed control room design review is expected to take about a year. Therefore, the Office of Nuclear Reactor Regulation (NRR) requires that those applicants for operating licenses who are unable to complete this review prior to issuance of a license make preliminary assessments of their control rooms to identify significant human factors and instrumentation problems and establish a schedule approved by NRC for correcting deficiencies. These applicants will be required to complete the more detailed control room reviews on the same schedule as licensees with operating plants.
Response
In accordance with the requirements of NUREG 0737 Supplement 1 LP&L performed a Detailed Control Room Design Review (DCRDR) in 1984-85. The DCRDR results and proposed changes to address human engineering discrepancies were submitted to the NRC via W3P85-1015 dated April 30, 1985 and supplemented via W3P86-2557 dated October 14, 1986 and W3P88-1240 dated August 3, 1988. By letter dated June 13, 1989, the NRC submitted a Supplemental Safety Evaluation which concluded that LP&L meets all of the nine DCRDR requirements of Supplement 1 to NUREG-0737.
1.9.16 PLANT SAFETY PARAMETER DISPLAY SYSTEM (I.D.2)
Position
Each applicant and licensee shall install a Safety Parameter Display System (SPDS) that will display to operating personnel a minimum set of parameters which define the safety status of the plant. This can be attained through continuous indication of direct and derived variables as necessary to assess plant safety status.
Response
Based on the requirement set forth in NUREG-0737, Supplement 1, the Waterford 3 SPDS has been designed to provide a concise display of critical plant variables to the control room operators to aid them in rapidly and reliably determining the safety status of the plant. Since the SPDS is a software implementation on the preexisting Plant Monitor Computer (PMC), no additional hardware, save dedicated SPDS terminals, was necessary.
Based on a review of the Waterford 3 FSAR, SER and draft Technical Specifications, it was determined that the implementation of the SPDS would have no adverse impact on the safe operation of existing instrumentation and equipment. Furthermore, the addition of the SPDS did not affect any FSAR analyses or Technical Specifications.
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A Safety Analysis Report of the Waterford 3 SPDS was developed in April, 1984 in order to provide the details of the SPDS implementation and to satisfy the NUREG-0737, Supplement 1 request for a written safety analysis. Included as part of the Waterford 3 SPDS Safety Analysis Report is a description of the PMC hardware and software (to include SPDS), the details of the SPDS Parameter Selection, the hierarchy of the SPDS displays, a discussion of the PMC reliability, a description of the Human Factors Principles employed, the implementation of the SPDS Verification and Validation as accomplished through the PMC Startup Testing and a summary of the Waterford 3 SPDS compliance with NUREG- 0737, Supplement 1.
The SPDS is further described in Appendix 7.7A.
1.9.17 TRAINING DURING LOW-POWER TESTING (I.G.1)
Position
Define and commit to a special low-power testing program approved by NRC to be conducted at power levels no greater than five percent for the purposes of providing meaningful technical information beyond that obtained in the normal startup test program and to provide supplemental training. (LBDCR 13-014, R309) Response
Chapter 14 (Section 14.2) reflects the startup and low power tests that were conducted for training. Subsection 13.2.1.1 describes the involvement of plant staff personnel with the low power test training. (LBDCR 13-014, R309) 1.9.18 REACTOR COOLANT SYSTEM VENTS (II.B.1)
Position
Each applicant and licensee shall install reactor coolant system (RCS) and reactor vessel head high point vents remotely operated from the control room. Although the purpose of the system is to vent noncondensible gases from the RCS which may inhibit core cooling during natural circulation, the vents must not lead to an unacceptable increase in the probability of a loss-of-coolant accident (LOCA) or a challenge to containment integrity. Since these vents form a part of the reactor coolant pressure boundary, the design of the vents shall conform to the requirements of Appendix A to 10CFR Part 50, "General Design Criteria." The vent system shall be designed with sufficient redundancy that assures a low probability of inadvertent or irreversible actuation.
Each licensee shall provide the following information concerning the design and operation of the high point vent system:
(1) Submit a description of the design, location, size, and power supply for the vent system along with results of analyses for loss-of-coolant accidents initiated by a break in the vent pipe. The results of the analyses should demonstrate compliance with the acceptance criteria of 10CFR50.46.
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(2) Submit procedures and supporting analysis for operator use of the vents that also include the information available to the operator for initiating or terminating vent usage.
Response
In accordance with the above position, Waterford 3 has included in its design a Reactor Coolant Gas Venting System to allow for remote venting of noncondensible gases, which may collect in the RCS, via a reactor vessel head vent or pressurizer steam space vent during post-accident situations. The design bases, a system description, and evaluation have been included in FSAR Subsection 5.4.15.
Test procedures were developed in accordance with subsection IWV of Section XI of the ASME Code for Category B valves.
Operating procedures were developed to address the use of the Reactor Coolant System Vents, defining the conditions under which the vents should be used or not used and including information pertaining to the initiating and terminating vent usage. Procedures for the use and non-use of RCS vents were prepared in accordance with the guidelines supplied by the NSSS vendor.
The vent system was reviewed as part of Item I.D.1 (see FSAR Subsection 1.9.15).
1.9.19 DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTAL QUALIFICATION OF EQUIPMENT FOR SPACES/SYSTEMS WHICH MAY BE USED IN POST-ACCIDENT OPERATIONS (II.B.2)
Position
With the assumption of a post-accident release of radioactivity equivalent to that described in Regulatory Guides 1.3 and 1.4 (i.e., the equivalent of 50 percent of the core radioiodine, 100 percent of the core noble gas inventory, and one percent of the core solids are contained in the primary coolant), each licensee shall perform a radiation and shielding-design review of the spaces around systems that may, as a result of an accident, contain highly radioactive materials. The design review should identify the location of vital areas and equipment, such as the control room, radwaste control stations, emergency power supplies, motor control centers, and instrument areas, in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during post-accident operations of these systems.
Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielding, or post-accident procedural controls. The design review shall determine which types of corrective actions are needed for vital areas throughout the facility.
Response
Refer to FSAR Appendix 12.3A
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1.9.20 POST-ACCIDENT SAMPLING CAPABILITY (II.B.3)
Position
A design and operational review of the reactor coolant and containment atmosphere sampling line systems shall be performed to determine the capability of personnel to promptly obtain (less than one hour) a sample under accident conditions without incurring a radiation exposure to any individual in excess of 3 and 18-3/4 rem to the whole body or extremities, respectively. Accident conditions should assume a Regulatory Guide 1.3 or 1.4 release of fission products. If the review indicates that personnel could not promptly and safely obtain the samples, additional design features or shielding should be provided to meet the criteria.
A design and operational review of the radiological spectrum analysis facilities shall be performed to determine the capability to promptly quantify (in less than two hours) certain radionuclides that are indicators of the degree of core damage. Such radionuclides are noble gas (which indicate cladding failure), iodines and cesiums (which indicate high fuel temperatures), and nonvolatile isotopes (which indicate fuel melting). The initial reactor coolant spectrum should correspond to a Regulatory Guide 1.3 or 1.4 release. The review should also consider the effects of direct radiation from piping and components in the auxiliary building and possible contamination and direct radiation from airborne effluents. If the review indicates that the analyses required cannot be performed in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet the criteria.
In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions. Procedures shall be provided to perform boron and chloride chemical analyses assuming a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4 source term). Both analyses shall be capable of being completed promptly (i.e., the boron sample analysis within an hour and the chloride sample analysis within a shift).
Response
Waterford 3 complies with this requirement as clarified in NUREG 0737 October 31, 1980 (see Subsection 9.3.8).
1.9.21 TRAINING FOR MITIGATING CORE DAMAGE (II.B.4)
Position
Licensees are required to develop a training program to teach the use of installed equipment and systems to control or mitigate accidents in which the core is severely damaged. They must then implement the training program.
Response
Comprehensive training in the use of installed plant systems to control or mitigate an accident in which the core is severely damaged is conducted during licensed operator training and requalification training. Subsection 13.2.2.4 reflects this commitment.
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1.9.22 PERFORMANCE TESTING OF BOILING-WATER REACTOR AND PRESSURIZED-WATER REACTOR RELIEF AND SAFETY VALVES (II.D.1)
Position
Pressurized-water reactor and boiling-water reactor licensees and applicants shall conduct testing to qualify the reactor coolant system relief and safety valves under expected operating conditions for design basis transients and accidents.
Response
The compliance of the Waterford 3 plant with NUREG 0737 Item II.D.1 requirements is described in W3P82-4011, dated 12/29/82. NRC acceptance of Waterford 3 compliance with this item is documented in a Safety Evaluation Report dated 04/15/88. Also refer to Subsection 3.9.3.3.
Waterford 3 design does not have block valves.
1.9.23 DIRECT INDICATION OF RELIEF AND SAFETY-VALVE POSITION (II.D.3)
Position
Reactor coolant system relief and safety valves shall be provided with a positive indication in the control room derived from a reliable valve-position detection device or a reliable indication of flow in the discharge pipe.
Response
To provide the operator with an unambiguous indication of primary safety valve position, a non-safety grade acoustic monitoring system was installed. The following discussion provides the manner in which the requirements of NUREG 0737 Item II.D.3 has been met.
The system ensures high reliability and testability and discriminates against inadvertent actuation due to impact events, cross talk and background noise. Two acoustic sensors in two separate instrumentation channels (one is mandated in paragraph (3) of NUREG 0737/II.D.3) are installed downstream of and close to each pressurizer safety valve. Their qualification for radiation and temperature exceeds actual DBA requirements. A metal enclosure protects the sensors against mechanical damage and provides support for coaxial cable in flexible conduit. The enclosure is installed on a machined mounting block strapped to the pipe. Any failure of the mounting bracket, however, does not compromise the function of either the safety valve or any other safety-related component.
The coaxial cable from the sensors to the containment penetrations is qualified for DBA conditions and/or suitably protected.
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The four converters/amplifiers are located in the RAB in a non-harsh environment area served by a safety-related HVAC system (SVS). The converted signal is processed and analyzed by four individual signal conditioners located in the Valve and Loose Parts Monitoring (V&LPM) Panel in the Control Room (see Subsection 4.4.6.1). The signal from all four monitors are combined into a single "open" position valve alarm on the RTG Board. In the lower part of this board, there is a pair of "open-closed" light indicators which enables the operator to determine which safety valve has opened.
The V&LPM Panel and converters/amplifiers are energized by a vital power source.
The utilization of information provided to the operator by this position monitoring system shall be integrated into alarm response and emergency procedures and into operator training.
Backup methods of determining valve position are provided which utilize each safety valve discharge line temperature and quench tank temperature and water level, as discussed in Subsection 5.2.5.1.3.
1.9.24 AUXILIARY FEEDWATER SYSTEM EVALUATION (II.E.1.1)
Position
The Office of Nuclear Reactor Regulation is requiring reevaluation of the auxiliary feedwater (AFW) systems for all PWR operating plant licensees and operating license applications. This action includes:
(1) Perform a simplified AFW system reliability analysis that uses event-tree and fault-tree logic techniques to determine the potential for AFW system failure under various loss-of-main-feedwater- transient conditions. Particular emphasis is given to determining potential failures that could result from human errors, common causes, single-point vulnerabilities, and test and maintenance outages;
(2) Perform a deterministic review of the AFW system using the acceptance criteria of Standard Review Plan Section 10.4.9 and associated Branch Technical Position ASB 10-1 as principal guidance; and
(3) Reevaluate the AFW system flowrate design bases and criteria.
Response
The Emergency Feedwater System requirements evaluation and reliability analysis appears as FSAR Appendices 10.4.9A and 10.4.9B.
(EC-33720, R307) Reference to Appendix 10.4.9B. Response to Position (1) appears as FSAR Appendix 10.4.9B. (EC-33720, R307)
Response to Positions (2) and (3) appear in table form in FSAR Appendix 10.4.9A as Tables 10.4.9A-1 and 10.4.9A-3 respectively.
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Additionally, a review of the Waterford 3 EFS design Technical Specifications and Operating Procedures against generic short term and long term recommendations is provided in FSAR Appendix 10.4.9A as Table 10.4.9A-2.
1.9.25 AUXILIARY FEEDWATER SYSTEM AUTOMATIC INITIATION AND FLOW INDICATION (II.E.1.2)
PART 1: Auxiliary Feedwater System Automatic Initiation-
Position
Consistent with satisfying the requirements of General Design Criterion 20 of Appendix A to 10 CFR Part 50 with respect to the timely initiation of the auxiliary feedwater system (AFWS), the following requirements shall be implemented in the short term:
(1) The design shall provide for the automatic initiation of the AFWS.
(2) The automatic initiation signals and circuits shall be designed so that a single failure will not result in the loss of AFWS function.
(3) Testability of initiating signals and circuits shall be a feature of the design.
(4) The initiating signals and circuits shall be powered from the emergency buses.
(5) Manual capability to initiate the AFWS from the control room shall be retained and shall be implemented so that a single failure in the manual circuits will not result in the loss of system function.
(6) The ac motor-driven pumps and valves in the AFWS shall be included in the automatic actuation (simultaneous and/or sequential) of the loads onto the emergency buses.
(7) The automatic initiating signals and circuits shall be designed so that their failure will not result in the loss of manual capability to initiate the AFWS from the control room.
In the long term, the automatic initiation signals and circuits shall be upgraded in accordance with safety- grade requirements.
PART 2: Auxiliary-Feedwater System Flowrate Indication
Position
Consistent with satisfying the requirements set forth in General Design Criterion 13 to provide the capability in the control room to ascertain the actual performance of the AFWS when it is called to perform its intended function, the following requirements shall be implemented:
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(1) Safety-grade indication of auxiliary feedwater flow to each steam generator shall be provided in the control room.
(2) The auxiliary feedwater flow instrument channels shall be powered from the emergency buses consistent with satisfying the emergency power diversity requirements of the auxiliary feedwater system set forth in Auxiliary Systems Branch Technical Position 10-1 of the Standard Review Plan, Section 10.4.9.
Response
Waterford 3's Emergency Feedwater System, as described in the referenced sections (see Table 1.9-1), complies with the requirements of Part I of this position.
Safety grade Emergency Feedwater flow indication and steam generator level (narrow range) indication is available to the operator in the control room. These instrument loops are powered by the uninterruptible 120V ac Class IE power system.
Wide range steam generator level indication is provided to complete compliance to Part II of this position. Each steam generator of the Waterford 3 plant is furnished with two Class IE redundant channels for wide range level indication. Each system has a range of 0 - 40 ft. water column and is displayed in the control room. The level sensors are used to develop control signals for the Emergency Feedwater System.
1.9.26 EMERGENCY POWER SUPPLY FOR PRESSURIZER HEATERS (II.E.3.1)
Position
Consistent with satisfying the requirements of General Design Criteria 10, 14, 15, 17 and 20 of Appendix A to 10 CFR Part 50 for the event of loss of offsite power, the following positions shall be implemented:
(1) The pressurizer heater power supply design shall provide the capability to supply, from either the offsite power source or the emergency power source (when offsite power is not available), a predetermined number of pressurizer heaters and associated controls necessary to establish and maintain natural circulation at hot standby conditions. The required heaters and their controls shall be connected to the emergency buses in a manner that will provide redundant power supply capability.
(2) Procedures and training shall be established to make the operator aware of when and how the required pressurizer heaters shall be connected to the emergency buses. If required, the procedures shall identify under what conditions selected emergency loads can be shed from the emergency power source to provide sufficient capacity for the connection of the pressurizer heaters.
(3) The time required to accomplish the connection of the preselected pressurizer heater to the emergency buses shall be consistent with the timely initiation and maintenance of natural circulation conditions.
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(4) Pressurizer heater motive and control power interfaces with the emergency buses shall be accomplished through devices that have been qualified in accordance with safety-grade requirements.
Response (LBDCR 15-028, R308A) Consistent with satisfying the requirements of General Design Criteria 10, 14, 15, 17, and 20 of Appendix A to 10CFR50 for the event of loss of offsite power, the Waterford 3 pressurizer heater power supply design provides the capability to supply, from either the offsite power source or the emergency power source (when offsite power is not available), a redundant group of pressurizer proportional heaters and associated controls necessary to establish and maintain natural circulation at hot standby conditions. Each group of heaters has access to only one Class IE division power supply. The Class IE interfaces for main power and control power are protected by safety-grade circuit breakers. Part of the closing circuitry to these safety-grade circuit breakers share a specific common circuit breaker, CVCEBKR014AB-13. If CVCEBKR014AB-13 is Open at the onset of a loss of offsite power, local manual operator action in the respective train switchgear room is necessary to reenergize the Pressurizer Heaters of that train. Being non-Class IE loads, the pressurizer heaters are automatically shed from the emergency power source upon the occurrence of a safety injection actuation signal. See FSAR Subsection 5.4.10.2 for a more detailed discussion. FSAR Section 8.3 has been revised to reflect this design. Figure 8.3-33 depicts the schematic arrangement of the emergency power supply for the pressurizer heaters. (LBDCR 15-028, R308A)
Emergency Procedures and training have been developed and implemented that provide information to the operator to ensure that: a) The diesel generator is not overloaded upon manually loading the heaters and b) The pressurizer heaters are available for pressure control to provide maintenance of natural circulation.
Technical Specifications reflect operability requirements for the pressurizer heaters.
1.9.27 DEDICATED HYDROGEN PENETRATIONS (II.E.4.1)
Position
Plants using external recombiners or purge systems for postaccident combustible gas control of the containment atmosphere should provide containment penetration systems for external recombiner or purge systems that are dedicated to that service only, that meet the redundancy and single-failure requirements of General Design Criteria 54 and 56 of Appendix A to 10CFR50, and that are sized to satisfy the flow requirements of the recombiner or purge system.
The procedures for the use of combustible gas control systems following an accident that results in a degraded core and release of radioactivity to the containment must be reviewed and revised, if necessary.
Response
This position is not applicable to Waterford 3. Waterford 3 has redundant recombiners permanently installed inside containment (see Table 1.9-1).
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Plant procedures for the use of combustible gas control systems following an accident resulting in a degraded core and release of radioactivity to the containment were reviewed and revised as necessary.
For additional information relating to the use of the hydrogen recombiners, refer to FSAR Subsection 6.2.5.
1.9.28 CONTAINMENT ISOLATION DEPENDABILITY (II.E.4.2)
Position
(1) Containment isolation system designs shall comply with the recommendations of Standard Review Plan Section 6.2.4 (i.e., that there be diversity in the parameters sensed for the initiation of containment isolation).
(2) All plant personnel shall give careful consideration to the definition of essential and nonessential systems, identify each system determined to be essential, identify each system determined to be nonessential, describe the basis for selection of each essential system, modify their containment isolation designs accordingly, and report the results of the reevaluation to the NRC.
(3) All nonessential systems shall be automatically isolated by the containment isolation signal.
(4) The design of control systems for automatic containment isolation valves shall be such that resetting the isolation signal will not result in the automatic reopening of containment isolation valves. Reopening of containment isolation valves shall require deliberate operator action.
(5) The containment setpoint pressure that initiates containment isolation for nonessential penetrations must be reduced to the minimum compatible with normal operating conditions.
(6) Containment purge valves that do not satisfy the operability criteria set forth in Branch Technical Position CSB 6-4 or the Staff Interim Position of October 23, 1979 must be sealed closed as defined in SRP 6.2.4, item II.3.f during operational conditions 1, 2, 3, and 4. Furthermore, these valves must be verified to be closed at least every 31 days.
(7) Containment purge and vent isolation valves must close on a high radiation signal.
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Response
(1) All containment isolation valves receive actuation signals from diverse sensed parameters. FSAR Table 6.2-32 reflects compliance to this requirement. As indicated in the table all isolation valves receive actuation from one or more of the following signals:
(a) Containment Isolation Actuation Signal (CIAS) - This signal is actuated by either high containment pressure or low pressurizer pressure
(b) Safety Injection Actuation Signal (SIAS) - This signal is generated by different circuitry and relays than CIAS, but as CIAS, is actuated by either high containment pressure or low pressurizer pressure
(c) Main Steam Isolation Signal (MSIS) - This signal is generated by low steam generator pressure or high containment pressure
(d) Containment Purge Isolation Signal,(CPIS) - This signal is generated by high containment radiation. This signal is used in combination with CIAS to isolate the Containment Purge valves (see FSAR Table 6.2-32 items 10 and 11).
(2) FSAR Table 6.2-32 indicates whether each system receiving the above signals is considered essential or non-essential. Essential systems are those necessary to assure:
(a) the integrity of the reactor coolant pressure boundary
(b) the capability to shutdown the reactor and maintain it in a safe shutdown condition
(DRN 04-1619, R14) (c) the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline in 10CFR50.67. (DRN 04-1619, R14)
The capability to manually override the isolation signal and reopen containment isolation valves on those non-essential systems which may be desirable to operate after the accident has been provided.
This feature is noted as a footnote in the Actuation Signal column of FSAR Table 6.2-32 and is in place on the valves listed in Table 1.9-3.
A review of the classification of essential and non-essential systems against the requirements of Revision 2 of Regulatory Guide 1.141 shall be conducted upon its issuance.
(3) As indicated in FSAR Table 6.2-32, all non-essential systems shall be automatically isolated by one or more of the above signals.
(4) Resetting of the isolation signals discussed above will not result in automatic. reopening of any containment isolation valves. Reopening of these valves can only be effected by deliberate operator action after reset of the signal. Each valve must then be opened individually by the operator.
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(5) Containment Setpoint Pressure
The containment pressure trip setpoint and allowable value for initiating containment isolation have been derived using the explicit setpoint methodology. This methodology applies a statistical combination of the individual uncertainty components (instrument loop error, setpoint variance, instrument drift, etc.) to establish a total instrument channel uncertainty. The trip setpoint established, therefore, ensures sufficient margin between the technical specification limit and the nominal trip setpoint. At the same time the setpoint is high enough to minimize inadvertent actuation of containment isolation.
(6) Containment Purge Isolation Valve (DRN 01-758) An analysis has been performed to determine the operability of the containment purge valves. This analysis has shown that the valves are capable of closing against the most severe design basis accident flow conditions when the valve opening is limited to 52 degrees. Modifications have been made to limit purge valve opening. (DRN 01-758) In addition, the purge valve isolation signals are designed such that they cannot be locked, reset, or overridden.
The FSAR has been revised to reflect consistency with the changes to actuation signals for the containment isolation system indicated in FSAR Table 6.2-32.
(7) As indicated in FSAR Table 6.2-32, the Containment Purge Isolations Valves (Pens. Nos. 10 and 11) are automatically isolated on CPIS (high radiation).
Waterford 3 Plant Operating Procedures entitled "Loss of Coolant" and "High Airborne Activity" contain details of isolation initiation and subsequent action to be taken. Additionally, diverse isolation signals have been installed for the purpose of containment isolation.
1.9.29 ADDITIONAL ACCIDENT-MONITORING INSTRUMENTATION (II.F.1) NOBLE GAS EFFLUENT MONITOR
Position
Noble gas effluent monitors shall be installed with an extended range designed to function during accident conditions as well as during normal operating conditions. Multiple monitors are considered necessary to cover the ranges of interest. (DRN 01-758) (1) Noble gas effluent monitors with an upper range capacity of 105 Ci/cc (Xe-133) are considered to be practical and should be installed in all operating plants. (DRN 01-758)
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(DRN 01-758) (2) Noble gas effluent monitoring shall be provided for the total range of concentration extending from normal condition (as low as reasonably achievable (ALARA) concentrations to a maximum of 105 Ci/cc (Xe-133). Multiple monitors are considered to be necessary to cover the ranges of interest. The range capacity of individual monitors should overlap by a factor of ten. (DRN 01-758) SAMPLING AND ANALYSIS OF PLANT EFFLUENTS
Position
Because iodine gaseous effluent monitors for the accident condition are not considered to be practical at this time, capability for effluent monitoring of radioiodines for the accident condition shall be provided with sampling conducted by adsorption on charcoal or other media, followed by onsite laboratory analysis.
CONTAINMENT HIGH-RANGE RADIATION MONITOR
Position
In containment radiation-level monitors with a maximum range of 108 rad/hr shall be installed. A minimum of two such monitors that are physically separated shall be provided. Monitors shall be developed and qualified to function in an accident environment.
CONTAINMENT PRESSURE MONITOR
Position
A continuous indication of containment pressure shall be provided in the control room of each operating reactor. Measurement and indication capability shall include three times the design pressure of the containment for concrete, four times the design pressure for steel, and -5 psig for all containments.
CONTAINMENT WATER LEVEL MONITOR
Position
A continuous indication of containment water level shall be provided in the control room for all plants. A narrow range instrument shall be provided for PWRs and cover the range from the bottom to the top of the containment sump. A wide range instrument shall also be provided for PWRs and shall cover the range from the bottom of the containment, to the elevation equivalent to a 600,000 gallon capacity. For BWRS, a wide range instrument shall be provided and cover the range from the bottom to five feet above the normal water level of the suppression pool.
(DRN 01-758)
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CONTAINMENT HYDROGEN MONITOR (DRN 01-758) Position
A continuous indication of hydrogen concentration in the containment atmosphere shall be provided in the Control Room. Measurement capability shall be provided over the range of 0 to 10 percent hydrogen concentration under both positive and negative ambient pressure.
Response
Noble Gas Effluent Monitors/Containment High Range Radiation Monitor/Sampling and Analysis of Plant Effluent
The accident - radiation monitoring instrumentation enables plant operators to better follow the course of a major accident and thereby assist them in making decisions with reference to mitigating the effects of a major accident. The instrumentation consists of the following monitors:
(1) One (1) Plant Vent Stack Monitor
(2) One (1) Condenser Vacuum Pump Effluent Monitor
(3) One (1) Fuel Handling Bldg Emergency Exhaust Effluent Monitor
(4) Two (2) Main Steam Line Monitors
(5) Two (2) High Range Containment Monitors
Readout of all monitor items for all of the accident radiation monitors is available from the Radiation Monitoring System Computer Remote Console CRT and from separate control room readouts.
Methods for converting instrument readings to release rates for unit time, based on exhaust air flow and considering radionuclide spectrum distribution as a function of time after shutdown, are described in Waterford 3 plant procedures. These procedures, numbered EP-2-050 and EP-2-051 respectively, are entitled "Offsite Dose Assessment" and "Offsite Dose Assessment (computerized)."
Additionally, the two (2) High Range Containment Monitors have separate readout via safety-related remote display/control devices. These remote display/control devices are mounted on the existing safety-related radiation monitoring cabinet (CP-14) located in the Control Room.
Equipment associated with monitors (1) through (4) above are qualified environmentally to IEEE 323-1974, IEEE 344-1975 and NUREG 0588. The High Range Containment Monitoring System including the display devices is qualified to IEEE 323-1974 and IEEE 344-1975 and NUREG 0588. The above is in adherence to Regulatory Guide 1.97 Revision 3, and NUREG 0737 requirements. The source of power for monitors (1) through (3) above are from a non-Class IE interruptible 120V ac source; the High Range Containment radiation monitor are powered from the Class IE interruptible 120V ac source; the Main Steam Line radiation monitors are powered from a non-class IE uninterruptible 120V ac source.
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The detectors associated with monitors (1) through (5) have a primary calibration report supplied by the vendor. This calibration report establishes the linearity of the detector. Field calibration of these detectors involves using a single calibration source which when placed in the proper geometry relative to the detector verifies a single point on the detector calibration curve. The frequency of field calibration for the monitors is every 18 months.
Table 1.9-4 summarizes the accident-radiation monitoring instrumentation data required per NUREG 0737.
Individual monitor descriptions follow:
1. High Range Noble Gas Plant Vent Monitor
(DRN 03-2054, R14) In adherence to NUREG 0737 and Regulatory Guide 1.97 Rev 3 one (1) High Range Noble Gas Monitor is installed to supplement the range of the existing plant vent stack radiation monitor. The high range noble gas monitor was purchased from General Atomic Company. The particular model which was purchased is entitled "Wide-Range Gas Monitor". This high range noble gas monitor makes use of separate isokinetic nozzles for isokinetic sampling over a flow range of 13,000 SCFM (±20%). This high range noble gas plant vent stack monitor is normally operating. (DRN 03-2054, R14)
In order to assure that plant personnel have access to certain assemblies of the monitor (such as the particulate and iodine sample filters) during a highrange release condition, the monitor is divided into separate assemblies that are located in such a way as to minimize personnel exposure to the postulated high levels of radiation. Figure 1.9-3 is a block diagram of the monitor showing the various assemblies of the system and their interconnections.
There are five assemblies: (1) Isokinetic Nozzles; (2) Sample Conditioner; (3) Wide-Range Gas Detectors; (4) Electronics; and (5) Readouts. Each of these assemblies are described below. Skid assemblies (2) and (3) are of open design to allow access to parts and to allow cooling by natural convection. All plumbing and piping are stainless steel and all connections are leak tested prior to shipment. All electrical power needed is distributed from the Wide-Range Gas Detector assembly.
(1) Isokinetic Nozzles
Two sets of isokinetic nozzles are normally used - one for normal and one for high-range conditions. Isokinetic nozzles are used to ensure representative particulate and iodine grab samples (see below). One isokinetic nozzle is mounted inside the duct; the other is mounted inside the sample stream coming from the duct-mounted nozzle. This second nozzle permits drawing of a 0.06 cfm sample under high activity conditions. The normal isokinetic nozzles operate at 1.67 ft3/min, whereas the high-range isokinetic nozzles operate at 0.06 ft3/min to minimize activity buildup. Included in this assembly are flow rate tranducers that are connected to a microprocessor to facilitate isokinetic flow control. The location for the nozzle assembly in the effluent stack was chosen in accordance with ANSI N13.1.
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(2) Sample Conditioner
This skid assembly is located downstream from the isokinetic nozzles. Its purpose is to provide representative particulate and iodine grab samples for laboratory analysis (in accordance with NUREG 0737), and to prevent contamination of the gas monitor by filtering out large concentrations of radioiodines and particulates. Without the sample conditioner, the monitor would become contaminated and remain upscale even when actual radioactive gas levels decreased. To provide enough filtering material to contain the radioiodines and particulates for the duration of the measured period, special multiple filters are used. Filters on the high-activity flowpath have full 4 X solid lead shielding to minimize personnel exposure. Fast disconnect fittings are provided for the grab sample filters. Grab sample actuation and duration are non-safety RMS control room panel functions (see below). The sample conditioning assembly shall be accessible during a high-range condition to retrieve grab samples. The instrumentation will function in an environment based on the Design Basis Shielding envelope assumptions presented in Table H.F.1-2 of NUREG 0737. The radiation exposures of personnel retrieving the samples will not exceed GDC 19 criteria. Detailed analysis of these filters for particulates and iodines is provided for via the gamma spectroscopy system described in FSAR Subsection 1.9.38. Particulate and iodine filter efficiencies are typically 99 percent.
(3) Wide-Range Gas Detectors
This skid assembly contains the three radioactive gas detectors; this assembly also contains the necessary pumps, flow control valves, flowmeters, etc. Each detector has a solenoid-actuated, checksource to verify proper operation and is full 4X cast-lead shielded to reduce background effects. The 11 decades of Noble gas concentrations are monitored continuously by the three detectors with at least one decade overlap between ranges of the individual detectors. Table 1.9-4 shows the ranges of the three detectors for Xe- 133. The low-range detector utilizes a plastic scintillator, whereas the mid-range and high-range detector are solid-state (Cd Te). As above, there are two flow paths through the detectors. During normal operation only the low-range detector is used and the mid-range and high-range detectors are bypassed. As the low-range detector begins to saturate, the flow path is automatically changed to the mid and high- range detectors and the low-range detector is purged. This prevents contamination of the low-range detector so that it will be available when it is automatically returned to service to measure radioactive gas concentrations as they return to low levels. The only solid-state electronics mounted on the skid are the detector preamplifiers which are provided with full 4X lead shielding to minimize radiation exposure. Without shielding, these electronics could not survive accident levels of background radiation for the duration of the accident condition.
(4) Electronics
The monitor is controlled by a microprocessor. The microprocessor performs flow control, valve actuations, engineering conversions, and other calculations and control functions, in addition to data storage. It is remotely located from the detectors, in a low radiation area. It contains the microprocessor, memory, high-voltage power supplies, preamplifiers, battery backup, etc. Mounted adjacent the microprocessor is a junction box for termination of user cables between the RM-80 and other assemblies of the monitor.
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(5) Readouts (DRN 01-758) The readout is from the Computer Remote Console CRT (CP-6) and from CP-14, in the Control Room. It is microprocessor based and provides a display of all monitored parameters. These include channel activity in Ci/cc flow rates, alarm status, etc. The local microprocessor also maintains history files of twenty-four 10- min, twenty-four 1-hour and twenty-eight 1-day averages of 4 activity that are available for recall via the CRT. Purge and grab sample control as well as effluent activity recording shall be provided for via labeled control switches and recorders, which will be located in CP-52 control cabinet. From the CP-52 cabinet, the operator is able to select clean prefilters, should one set be loaded to the points where appreciable concentrations of xenon off-gas is produced by iodine decay. In this manner, the Noble gas detector will not interpret iodine-daughter xenon as Noble gas going out the stack. Additionally from this same cabinet the operator may also take grab samples of 1 to 99 minutes duration for the low-range flow path and 0 to 99 seconds for the high-range flow path. (DRN 01-758) 2. Condenser Vacuum Pumps Effluent Monitor
This monitor is installed to monitor noble gas effluents from the condenser vacuum pump. Its operation shall be as described for the high-range plant vent stack monitor. Notable exception will be that it shall operate at all times that the condenser vacuum pumps are operating. Due to the high humidity conditions of the sample stream, a large amount of particulate plateout is expected in the sample lines of this monitor. As a result of this expected high plateout, representative samples cannot be assured even with isokinetic sampling. Thus there are no provisions made for isokinetic sampling for this monitor. Filters are used to prevent particulate and iodine activity buildup in the Noble gas monitor. Additional details on the functions provided by this monitor are given in FSAR Subsection 11.5.2.4.1.5.
3. Fuel Handling Building Emergency Exhaust Effluent Monitor
A monitor is installed to monitor the Fuel Handing Building Emergency Exhaust Effluent path. It is of the same type as the Plant Vent Stack monitor.
The operation of the FHB Emergency Exhaust Effluent Monitor shall be as follows:
a) Upon a high radiation signal the potentially contaminated area of the Fuel Handling Building shall be automatically isolated to prevent an unrestricted release of airborne effluents from occurring (see Subsection 9.4.2). This isolation signal shall cause both emergency exhausts to begin operating.
b) The Fuel Handling Building Emergency Exhaust Effluent Monitor shall begin operating at the same time that the Fuel Handling Building emergency exhaust begins operation. The monitor at this time shall be drawing a sample from both emergency exhausts via isokinetic nozzles located inside each emergency exhaust duct. Sampling at this time shall not be isokinetic.
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c) Control Room operators after the automatic startup of both emergency exhausts shall select only one emergency exhaust for further use.
d) Selection of one emergency exhaust shall cause a solenoid operated valve to isolate the sample stream coming from the isokinetic nozzle located in the other FHB emergency exhaust. Thus, the monitor shall now be drawing an isokinetic sample from the operator selected FHB emergency exhaust duct.
e) Failure in one emergency exhaust system shall cause the isokinetic nozzle sample stream isolation valve associated with that emergency exhaust system to close, thereby assuring true isokinetic sampling from the other emergency exhaust.
The sequence described above shall be dependent only upon operator selection of one FHB emergency exhaust system. Other monitor functions shall be as described for the plant vent stack monitor.
4. Main Steam Line Monitors
In order to estimate the releases which may occur as a result of the actuation of steam generator secondary relief valves (SRV) and atmospheric steam dump valves (ADV) one collimated GM tube is installed to view the activity of each main steam line. The monitors are mounted within a three inch thick lead shield with a "window" at the front of the detector. The detector reading is in mr/hr, and shall be recorded by the microprocessor as a minimum of twenty-four 10 min, 24 -one hour and 28 -one day averages. (DRN 01-758) Calculational methods are employed to quantify radiological releases based on monitor dose rates. In order to obtain concentrations of 10-1 to 103 Ci/cc of Xe-133 in the main steam line a large primary to secondary leak must be present coincident with a large amount of cladding failure. Present with Xe-133 will be other nuclides. Based on the postulated primary system accident scenarios (e.g., cladding failure, fuel failure) and on the assumed steam generator isotope partition factors, isotopic concentrations in the main steam lines can be calculated. Two conversion factors have been developed for the main steam line monitors in terms of mr/hr per Ci/cc of pressurized steam. These factors are based on isotopic fractions arising from gross fuel failure and one percent cladding failure, respectively. A decontamination factor (DF) of 100 is assumed for iodine passing through the steam generators. Noble gases are assumed to pass through the steam generators unhindered. All the applicable noble gases and iodines were considered in the conversion factors. An average energy per disintegration per isotope was used to determine the dose-rate, and the primary dose rate contributors were Kr-87 and Kr-88. The methodology used to determine the conversion factors is taken from the Reactor Shielding Design Manual, 1956, by T Rockwell III. This model accounts for the thickness of the main steam line wall. (DRN 01-758) Estimating radiological releases through the SRV's and the ADV's is calculated by summing the products of the a) concentrations and b) the mass of steam released. Main steam concentrations are obtained from a calculated conservative conversion factor. The mass of steam released can be monitored by using the Main Steam Flow Recorder located between the steam generator and the main steam safety valves.
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5. High Range Containment Monitors
The High-Range Radiation Monitor consists of a gamma detector (General Atomic model number RD-23) and cable suitable for use in a containment environment, and support electronics including a readout located in the Control Room. The detector is encased in stainless steel to protect it from containment sprays and high temperatures. The monitor is a safety monitor (Class 1E) and is qualified under LOCA conditions to IEEE 323-1974. Radiation levels of up to 108 R/hr are displayed in the control room on a front panel meter. Two level trips are provided for alert and high radiation levels and are independently adjustable over the full range. A failure trip is provided to actuate upon loss of power, high voltage, or signal from the detector. Automatic self-testing is provided to continuously verify detector operation. Outputs are provided for recorders, remote alarm relays, and meters. A separate local display of radiation levels and alarm conditions exists. Energy response is uniform (± 20 percent) for photons in the range of 80 Kev to 3 Mev.
Readout control and data recording for these monitors is provided for as described for Class IE area monitors in FSAR Subsection 12.3.4. The monitors shall be located in containment in a manner as to provide a reasonable assessment of area radiation conditions inside containment. The monitors shall be widely separated so as to provide independent measurements and shall "view" a large fraction of containment volume. Monitors shall not be placed in areas which are protected by massive shielding and shall be reasonably accessible for replacement, maintenance and calibration.
A sustaining signal is generated within the detector corresponding to a predetermined value. A failure alarm will occur if the signal from the detector falls below this value. This feature assures knowledge of the monitor's integrity at all times.
Containment Hydrogen Monitor
See FSAR Subsection 6.2.5.1.
Containment Pressure Monitor
Response
Waterford 3 complies with this requirement.
A continuous recording of containment wide range (-5 to +195 psig) pressure is provided in the control room. This recorded range is approximately four times the design pressure of Waterford 3's steel containment.
(EC-12329, R306) Containment wide range pressure monitoring instrumentation consists of 2 redundant Class IE channels. Each channel consists of a pressure transmitter, which utilizes penetration #54 and is physically located outside the containment in the Auxiliary Building, approximately at EL. +5 feet. The accuracy of the transmitter to be used is ± 5% of the span. The pressure transmitter output signal is processed by a process analog control system (PAC) which in turn furnishes signals for the recorder in the main control room and the plant monitoring computer. The entire range of -5 to +195 psig is recorded by one trace of the recorder. A visual indicator is part of the recorder. (EC-12329, R306)
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Qualification is in accordance with the criteria for Class IE transmitters located outside the containment building. The Containment pressure transmitters meet the requirements of Appendix B to NUREG-0737.
Containment Water Level Monitor
Response
Waterford 3 complies with this requirement.
Two redundant Class IE channels of instrumentation are provided to monitor containment sump level (narrow range). The narrow range monitors meet the recommendations of Regulatory Guide 1.89. Each channel of instrumentation consists of the following: (DRN 99-1035) a) A level transmitter with a range of 0 - 15', located inside the containment sump. The total depth of the containment sump is 14'. The measurement covers the range from 1.5' at the bottom of the sump to 16.5', thus, total approximate range is 0 to 15'. The containment sump will begin to overflow at an indicated level of 12.5'. (DRN 99-1035) b) A process analog control system (PAC) to monitor level transmittal signal and to develop output signals to the plant monitoring computer and recorder or indicator. c) A recorder/indicator, mounted on the main control board, for registering the containment sump level. One channel is provided with a recorder and one channel is provided with an indicator.
Two redundant Class IE channels of instrumentation are provided to monitor containment flood level (wide range). The wide range monitors meet the requirements of Appendix B to NUREG-0737. Each channel of instrumentation consists of the following: a) A level transmitter with an approximate range of 0 - 16' located inside the containment. The measurement covers the range from the floor level at elevation -14.8 feet to elevation + 1.4 feet. The maximum flood level established by calculation is within this range. b) A process analog control system (PAC) to monitor level transmittal signal and to develop output signals to the plant monitoring computer and recorder or indicator. c) A recorder/indicator, mounted on the main control board, for registering the containment flood level. One channel is provided with a recorder and one channel is provided with an indicator.
(DRN 99-1035) (DRN 99-1035)
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1.9.30 INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING (II.F.2)
Position
Licensees shall provide a description of any additional instrumentation or controls (primary or backup) proposed for the plant to supplement existing instrumentation (including primary coolant saturation monitors) in order to provide an unambiguous, easy-to-interpret indication of inadequate core cooling (ICC). A description of the functional design requirements for the system shall also be included. A description of the procedures to be used with the proposed equipment, the analysis used in developing these procedures, and a schedule for installing the equipment shall be provided.
Response
See Appendix 1.9A.
1.9.31 EMERGENCY POWER FOR PRESSURIZER EQUIPMENT (II.G.1)
Position
Consistent with satisfying the requirements of General Design Criteria 10, 14, 15, 17, and 20 of Appendix A to 10CFR Part 50 for the event of loss-of-offsite power, the following positions shall be implemented.
Power Supply for Pressurizer Relief and Block Valves and Pressurizer Level Indicators
(1) Motive and control components of the power-operated relief valves (PORVs) shall be capable of being supplied from either the offsite power source or the emergency power source when the offsite power is not available. (DRN 01-758) (2) Motive and control components associated with the PORV block valves shall be capable of being supplied from either the offsite power source or the emergency power source when the offsite power is not available. (DRN 01-758) (3) Motive and control power connections to the emergency buses for the PORVs and their associated block valves shall be through devices that have been qualified in accordance with safety-grade requirements.
(4) The pressurizer level indication instrument channels shall be powered from the vital instrument buses. The buses shall have the capability of being supplied from either the offsite power source or the emergency power source when offsite power is not available.
Response
(1) Waterford 3 has spring loaded pressurizer safety valves; therefore, this position requirement is not applicable.
(2) Waterford 3 has no pressurizer block valves; therefore, this position requirement is not applicable.
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(3) See (1) and (2) above.
(4) Waterford 3 pressurizer level indication instrument channels are powered from the Class IE uninterruptible 120V ac power system. This system can be supplied from either the offsite power source or the standby power supply (Emergency Diesel Generators). (See Table 1.9-1).
For CWDs and Electrical Schematics, see FSAR Table 1.7-1.
1.9.32 IE BULLETINS ON MEASURES TO MITIGATE SMALL-BREAK LOCAs AND LOSS OF FEEDWATER ACCIDENTS (II.K.1)
Position
Review all valve positions, positioning requirements, positive controls and related test and maintenance procedures to assure proper ESF functioning. (C.1.5)
Response
A review of all ESF valve positions, controls and test and maintenance procedure is conducted during procedure preparation to ensure proper ESF functioning. This review assures normal lineup in the mode required for ESF operation. Test and maintenance procedures are reviewed so as to be "stand alone" procedures with independent verification for valve positioning and restoration.
Position
Review and modify, as required, procedures for removing safety-related systems from service (and restoring to service) to assure operability status is known. (C.1.10)
Response
Procedures dealing with all safety related systems and components require independent verification of all valve operations and breaker positions both for operational safeguards readiness and for realignment for the performance of maintenance or tests. At the completion of maintenance or tests, the same independent verification is performed for the restoration portion of the procedure to assure the operability status is known. Periodic tests and checks are performed to verify continued operability status.
1.9.33 ORDERS ON B&W PLANTS (II.K.2)
THERMAL MECHANICAL REPORT--EFFECT OF HIGH-PRESSURE INJECTION ON VESSEL INTEGRITY FOR SMALL-BREAK LOSS-OF- COOLANT ACCIDENT WITH NO AUXILIARY FEEDWATER (II.K.2.13)
Position
A detailed analysis shall be performed of the thermal-mechanical conditions in the reactor vessel during recovery from small breaks with an extended loss of all feedwater.
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Response
This analysis has been performed by the CE Owners Group (CEOG) and is discussed in CEN-189, submitted to the NRC by the CEOG on 12/31/81. Appendix I of this report specifically addresses Waterford 3.
POTENTIAL FOR VOIDING IN THE REACTOR COOLANT SYSTEM DURING TRANSIENTS (II.K.2.17)
Position
Analyze the potential for voiding the reactor coolant system (RCS) during anticipated transients.
Response
Louisiana Power & Light Company sponsored the CE Owners Group evaluation of this item. The evaluation results are contained in CEN-199, "Effects of Vessel Head Voiding During Transients and Accidents in CE NSSS's".
It was found that voiding in the reactor vessel upper head region is not expected to occur for normal operational transients. For natural circulation cooldown transients voiding may occur in the reactor vessel upper head region. However, in the event that voids are formed, the operator guidance provided in CE's emergency procedure guidelines adequately address how to control and reduce the voids. These guidelines form the basis for the emergency operating procedures at Waterford 3.
Additionally, for FSAR Chapter 15 transients the impact of voiding will not result in violation of the Standard Review Plan requirements. Finally, the report concludes that any potential void formation during the plant transients addressed is not great enough to impair reactor coolant circulation or core coolability.
SEQUENTIAL AUXILIARY FEEDWATER FLOW ANALYSIS (II.K.2.19)
Position
Provide a benchmark analysis of Sequential Auxiliary Feedwater (AFW) flow to the steam generators following a loss of main feedwater.
Response
The concerns expressed in this item, and as clarified in NUREG-0737, are not considered applicable to Waterford 3 which utilizes vertical U-tube steam generators as designed by Combustion Engineering.
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1.9.34 FINAL RECOMMENDATIONS OF B&O TASK FORCE (II.K.3)
INSTALLATION AND TESTING OF AUTOMATIC POWER-OPERATED RELIEF VALVE ISOLATION SYSTEM (II.K.3.1)
Position
All PWR licensees should provide a system that uses the PORV block valve to protect against a small- break loss-of-coolant accident. This system will automatically cause the block valve to close when the reactor coolant system pressure decays after the PORV has opened. Justification should be provided to assure that failure of this system would not decrease overall safety by aggravating plant transients and accidents.
Each licensee shall perform a confirmatory test of the automatic block valve closure system following installation.
Response
The requirements of this position are not applicable to Waterford 3. Waterford 3 has no pressurizer block valves.
REPORT ON OVERALL SAFETY EFFECT OF POWER-OPERATED RELIEF VALVE ISOLATION SYSTEM (II.K.3.2)
Position
(1) The licensee should submit a report for staff review documenting the various actions taken to decrease the probability of a small-break loss-of-coolant accident (LOCA) caused by a stuck-open power-operated relief valve (PORV) and show how those actions constitute sufficient improvements in reactor safety.
(2) Safety-valve failure rates based on past history of the operating plants designed by the specific nuclear steam supply system (NSSS) vendor should be included in the report submitted in response to (1) above.
Response
Waterford 3 has spring loaded pressurizer safety valves; therefore, the requirements of this position are not applicable to Waterford 3.
REPORTING OF SV AND RV FAILURES AND CHALLENGES (II.K.3.3)
Position
Assure that any failure of a PORV or safety valve to close will be reported to the NRC promptly. All challenges to the PORVs or safety valves should be documented in the annual report.
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Response
In the unlikely event of any failure of the spring loaded pressurizer safety valves to close, LP&L will report such failures to the NRC promptly. All challenges to the safety valves will be documented in the monthly report.
AUTOMATIC TRIP OF REACTOR COOLANT PUMPS DURING LOSS-OF-COOLANT ACCIDENT (II.K.3.5)
Position
Tripping of the reactor coolant pumps in case of a loss-of-coolant accident (LOCA) is not an ideal solution. Licensees should consider other solutions to the small-break LOCA problem (for example, an increase in safety injection flow rate). In the meantime, until a better solution is found, the reactor coolant pumps should be tripped automatically in case of a small-break LOCA. The signals designated to initiate the pump trip are discussed in NUREG-0623.
Response
LP&L has been a charter participant in the CE Owners Group development of CEN-152, CE Emergency Procedure Guidelines (EPGs). The EPGs form the basis, as approved by the NRC, for the Waterford 3 emergency operating procedures (EOPs). For Cycle 2 operation, LP&L implemented Revision 2 to CEN- 152.
EVALUATION OF POWER-OPERATED RELIEF VALVE OPENING PROBABILITY DURING OVERPRESSURE TRANSIENT (II.K.3.7)
Position
Most overpressure transients should not result in the opening of the power-operated relief valve (PORV). Therefore, licensees should document that the PORV will open in less than 5 percent of all anticipated overpressure transients using the revised setpoints and anticipatory trips for the range of plant conditions which might occur during a fuel cycle.
Response
Waterford 3 has spring loaded pressurizer safety valves; therefore, the requirements of this position are not applicable to Waterford 3.
(DRN 01-758)
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REPORT ON OUTAGES OF EMERGENCY CORE-COOLING SYSTEMS LICENSEE REPORT AND PROPOSED TECHNICAL SPECIFICATION CHANGES (II.K.3.17) (DRN 01-758) Position
Several components of the emergency core-cooling (ECC) systems are permitted by technical specifications to have substantial outage times (e.g., 72 hours for one diesel-generator; 14 days for the HPCI system). In addition, there are no cumulative outage time limitations for ECC systems. Licensees should submit a report detailing outage dates and lengths of outages for all ECC systems for the last five years of operation. The report should also include the causes of the outages (i.e., controller failure, spurious isolation).
Response
A formal reliability/availability program has been established for Waterford 3. The information will contain: 1) outage dates and duration of outages; 2) cause of the outage; 3) ECCS systems or components involved in the outage; and 4) corrective action taken. Requirements for collecting and analyzing data specific to the ECCS as well as many other plant systems is incorporated in the procedures that define the reliability/availability program. Data collection and analysis is integrated and interfaced with that required for LER and NPRDS reporting. The methodology is designed to provide a means for quick retrieval of ECCS information. (DRN 01-758) By letter dated May 5, 1989, the NRC informed LP&L that Item II.K.3.17 of NUREG-0737 was an early action item to allow the NRC to quickly evaluate existing requirements and prepare followup actions. The requirements of 10CFR50.72 and industry efforts to report on the Equipment Performance Information and Exchange System (EPIX) are adequate for reporting ECCS outages. A special report from Waterford 3 is not required. Item II.K.3.17 is closed for Waterford 3. (DRN 01-758) EFFECT OF LOSS OF ALTERNATING-CURRENT POWER ON PUMP SEALS (II.K.3.25)
Position
The licensees should determine, on a plant-specific basis, by analysis or experiment, the consequences of a loss of cooling water to the reactor recirculation pump seal coolers. The pump seals should be designed to withstand a complete loss of alternating-current (ac) power for at least two hours. Adequacy of the seal design should be demonstrated.
Response
Waterford 3 complies with this position by supplying emergency power (Diesel Generator automatic loading) to the component cooling water pump. Although the component cooling water lines to the reactor recirculation pump seal coolers are isolated on a CSAS (See FSAR Table 6.2-32), these lines and isolation valves are provided with a manual override.
(DRN 01-758)
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REVISED SMALL-BREAK LOSS-OF-COOLANT-ACCIDENT METHODS TO SHOW COMPLIANCE WITH 10CFR PART 50, APPENDIX K (II.K.3.30) (DRN 01-758) Position
The analysis methods used by nuclear steam supply system (NSSS) vendors and/or fuel suppliers for small-break loss-of-coolant accident (LOCA) analysis for compliance with Appendix K to 10CFR Part 50 should be revised, documented, and submitted for NRC approval. The revisions should account for comparisons with experimental data, including data from the LOFT Test and Semiscale Test facilities.
Response
LP&L has participated in a series of CE Owner's Group tasks in support of Item II.K.3.30.
In the summer of 1979, CE submitted two reports to the NRC, CEN-114-P, "Review of Small Break Transients in Combustion Engineering Nuclear Steam Supply Systems," July 1979 (Proprietary), and CEN- 115-P, "Response to NRC IE Bulletin 79-06C, Items 2 and 3 for CE Nuclear Steam Supply Systems," August 1979 (Proprietary), which describe CE's Small Break LOCA Evaluation Model. These submittals were prepared in response to NRC requests following the TMI-2 accident. After review of these documents, the NRC identified a number of questions with some portions of the small break model. The NRC requested a response to these questions in the NRC TMI Action Plan, NUREG-0737, Item II-K-3.30. At a meeting held on January 26, 1981 with members of the NRC staff and representatives of the CE Owners Group and CE, the NRC staff described seven technical items which form the basis for the seven specific questions relative to the CE Small Break LOCA Evaluation Model. The NRC staff also indicated that responding to these seven questions would fulfill the response to Item II.K.3.30 of the NRC TMI Action Plan.
The seven questions were responded to in CEN-203-P Revision I-P, "Response to NRC Action Plan Item II.K.3.30 Justification of Small Break LOCA Methods," transmitted to the NRC by CE on April 15, 1982. The response shows that using the CE Small Break LOCA Evaluation Model results in-conservatively high cladding temperatures, and complies with Waterford 3's requirement under this item.
By letter dated July 12, 1985 the NRC found LP&L in compliance with the requirements of Item II.K.3.30 and that a plant specific analysis (Item II.K.3.31) was not required.
PLANT-SPECIFIC CALCULATIONS TO SHOW COMPLIANCE WITH 10CFR PART 50.46, (II.K.3.31)
Position
Plant-specific calculations using NRC-approved models for small-break loss-of-coolant accidents (LOCAS) to show compliance with 10CFR50.46 should be submitted for NRC approval by all licensees.
Response
See Response to Item II.K.3.30 above.
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1.9.35 EMERGENCY PREPAREDNESS-SHORT TERM (III.A.1.1)
Position
Comply with Appendix E, "Emergency Facilities," to 10CFR Part 50, Regulatory Guide 1.101, "Emergency Planning for Nuclear Power Plants," and for the offsite plans, meet essential elements of NUREG-75/111 or have a favorable finding from FEMA.
Provide an emergency response plan in substantial compliance with NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants" except that only a description of and completion schedule for the means for providing prompt notification to the population, the staffing for emergencies in addition to that already required, and an upgraded meteorological program need be provided. NRC will give substantial weight findings on offsite plans in judging the adequacy against NUREG-0654. Perform an emergency response exercise to test the integrated capability and a major portion of the basic elements existing within emergency preparedness plans and organizations.
Response
The Waterford 3 Emergency Plan was updated using the criteria provided in NUREG-0654 with special attention to the establishment of the emergency action levels in accordance with NUREG-0610. Refer to FSAR Section 13.3.
The Technical Support Center, the Operational Support Center, and the near site Emergency Operations Facility have been established. Refer to FSAR Section 13.3.
Improved offsite radiological monitoring capability in accordance with NRR/RAB technical position has been developed.
Louisiana Power & Light coordinated with the State and local government in developing Radiological Emergency Response Plans for Waterford 3. Therefore, the State, local and facility plans are well coordinated.
Periodic exercises are conducted with Federal, State, and local government to evaluate major portions of their emergency response capability and to correct identified deficiencies. Refer to FSAR Section 13.3.
1.9.36 UPGRADE EMERGENCY SUPPORT FACILITIES (III.A.1.2)
Position
Each operating nuclear power plant shall maintain an onsite technical support center (TSC) separate from and in close proximity to the control room that has the capability to display and transmit plant status to those individuals who are knowledgeable of and responsible for engineering and management support of reactor operations in the event of an accident. The center shall be habitable to the same degree as the control room for postulated accident conditions. The licensee shall revise his emergency plans as necessary to incorporate the role and location of the TSC. Records that pertain to the as-built conditions and layout of structures, systems, and components shall be readily available to personnel in the TSC.
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An operational support center (OSC) shall be established separate from the control room and other emergency response facilities as a place where operations support personnel can assemble and report in an emergency situation to receive instructions from the operating staff. Communications shall be provided between the OSC, TSC, EOF, and control room.
An emergency operations facility (EOF) will be operated by the licensee for continued evaluation and coordination of all licensee activities related to an emergency having or potentially having environmental consequences.
Response
The Technical Support Center, Operational Support Center, and the near site Emergency Operations Facility have been established. Refer to FSAR Section 13.3.
1.9.36a IMPROVING LICENSEE EMERGENCY PREPAREDNESS-LONG-TERM (III.A.2)
Position
Each nuclear facility shall upgrade its emergency plans to provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. Specific criteria to meet this requirement is delineated in NUREG-0654 (FEMA-REP-1), "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparation in Support of Nuclear Power Plants."
Response
The Emergency Plan was upgraded in accordance with the applicable criteria of NUREG-0654. Refer to FSAR Section 13.3.
1.9.37 INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT LIKELY TO CONTAIN RADIOACTIVE MATERIAL FOR PRESSURIZED-WATER REACTORS AND BOILING-WATER REACTORS (III.D.1.1)
Position
Applicants shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as-practical levels. This program shall include the following:
(1) Immediate leak reduction
(a) Implement all practical leak reduction measures for all systems that could carry radioactive fluid outside of containment.
(b) Measure actual leakage rates with system in operation and report them to the NRC.
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(2) Continuing Leak Reduction -- Establish and implement a program of preventive maintenance to reduce leakage to as-low-as-practical levels. This program shall include periodic integrated leak tests at intervals not to exceed each refueling cycle.
Response
LP&L has instituted a program to maintain leakage rates of systems outside containment which could contain radioactivity to as low as practical. To support this program, a review of plant systems has identified the systems outside containment which could potentially contain highly radioactive fluids following a serious accident.
A. Systems included in the leak reduction program:
1) Containment Spray System - that portion of the system located outside containment that would be in use in the recirculation mode of operation including that suction piping from the Safety Injection Sump up through the pumps and heat exchangers to the containment isolation valve.
2) Low Pressure Safety injection - the piping outside of containment in use during operation in the shutdown cooling mode.
3) High Pressure Safety Injection - the piping from the recirculation suction header through the pump up to containment.
4) Hydrogen Analyzer System - that portion of piping from the outside containment isolation valve to the Hydrogen Analyzer Panels, along with the piping to the containment atmospheric grab sampler in the post-accident sampling area and return piping back to the containment isolation valve. (DRN 01-758) 5) Post-accident Sampling System (PASS) - for the liquid portion of the system, testing includes the piping from the connection to the Primary Sample System sample point 5A and 5B sample lines to the PASS skid packages back to the outside containment isolation valve. Also included is the piping from the RCS Hot leg Sample outside containment isolation valve to the PASS skid package and back to the containment isolation valve. For the gas portion of the system, also included is that portion of the tubing from the liquid/gas separator through the skid package back to the outside containment isolation valve. (DRN 01-758) 6) Containment Vacuum Relief (CVR) - the essential instrument tubing from outside containment to the differential pressure instruments. (DRN 01-758) 7) Primary Sampling System - that portion of primary sample point 5A and 5B sampling lines from the safety injection recirculation lines to their connection to the Post Accident Sampling System line. (DRN 01-758) B. Systems excluded from the program (their isolation will not preclude any option of cooling the reactor core nor prevent the use of needed safety systems):
1) The Gaseous Waste Management System. This system isolates on CIAS and is not required for use post-accident. The Reactor Coolant Vent System provides RCS venting as discussed in Subsection 1.9.18.
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2) The Chemical Volume and Control System (Charging and Letdown). On a CIAS or SIAS this system will isolate letdown flow which is not required after an accident nor is it needed to bring the reactor to a safe shutdown condition. In addition to letdown being isolated on an SIAS, the Volume Control Tank (VCT) outlet valve closes resulting in the charging pumps taking suction directly from the Boric Acid Makeup Tanks, or the charging pumps can be lined up to the Refueling Water Storage Pool (RWSP). Thus, no highly radioactive fluids are expected to flow through the portion of the CVCS outside of containment.
3) Reactor Coolant Pump Seal Bleed-off to the VCT. This system is isolated on a CIAS. If seal bleed-off is needed post-accident, the pressure in that portion of the system inside containment will increase and the header relief valve will open thus providing a flow path to the Quench Tank.
4) The Boron Management System. This system receives a CIAS which isolates the Reactor Drain Tank outlet, thus, when the tank is pressurized it relieves to the containment sump.
5) The Primary Sampling System. This system with the exception of the portions of sample point 5A and 5B sampling lines discussed in A. 6) above, isolates on a CIAS and would not be required because of the availability of the Post-Accident Sampling System.
6) The Shield Building Ventilation System. That portion of the system from the annulus through the filters and up to the fan is operated at a negative pressure. So, any leakage would be in the inward direction and not outward from the system. System leakage downstream of the fan is of no radiological significance since the SBVS filter exhaust is suitable for discharge to the atmosphere. (DRN 04-1619, R14) 7) The Controlled Ventilation Area System. Similar to the Shield Building Ventilation System, that portion of the system up to the fan is operated at a negative pressure, and the discharge of the fan is of no radiological significance since the CVAS filter exhaust is suitable for discharge of the environment. (DRN 04-1619, R14) (DRN 05-1265, R14-A) C. For liquid systems, leakage detection is performed by visual inspection of all potential leak sources (e.g., valves, pump seals, etc.). Upon detection of a leak, the leak rate is determined. For gas systems, leakage detection is performed by pressurizing the system with an inert gas, nitrogen, or instrument air and visually inspecting potential leak sources with a soapy water solution (or equivalent method). Those leakage sources whose leak rates cannot be reduced to as low as practical, will be reported to the Plant Manager or his designee for resolution. Initial leakage rates were determined during plant startup testing prior to initial criticality and reported to the NRC in W3P85-0538 dated March 4, 1985. Future leak rate measurements will be performed at intervals not to exceed each refueling outage. Records of leakage rates and their sources will be retained in plant files. (DRN 05-1265, R14-A)
D. The potential release path identified in the NRC letter dated October 17, 1979 (Radioactive Release at North Anna Unit 1 and Lessons Learned) is not credible in the Waterford 3 design. High level in the volume control tank is alarmed in the main control room and automatically causes influent flow to be diverted to the Boron Management System. The overall program for prevention of unplanned radioactivity releases will incorporate the features of IE Circular 79-21. Aspects of the program and related features of the Waterford 3 design are:
1) All tanks outside of containment are provided with level indicators and high
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level alarms to alert the operator of high level conditions, and loop seals on overflow lines to prevent the escape of radioactive gas. Generally, collection tanks and tanks which receive processed waste are provided with backup tanks. Tanks outside of containment which are not provided with a backup tank are:
- Primary Water Storage Tank (PWST) - The PWST is located outside the nuclear plant island and as such, the Technical Specifications place a strict limit on the amount of activity allowed in the tank. Therefore, a spill from the tank would not involve a significant amount of radioactivity. (DRN 00-803) - Equipment Drain Tank (EDT) - The EDT does not have an overflow and thus there is no potential for spillage. In the event of high level, the EDT pump starts and pumps to the Holdup Tanks. (DRN 00-803) - Spent Resin Tank (SRT) - High level in the SRT automatically causes the inlet valve to close. The SRT is vented to the vent gas collection header. (DRN 01-758) - Condensate Storage Tank - Overflow from this tank is to the floor of its enclosed concrete area, which is provided with six inch curbs to limit the spread of liquid. (DRN 01-758) 2) Storm drains are located away from areas with a high potential for radioactive spills and there are no cross-connects between the floor and storm drainage systems.
3) Radioactive pumps are generally located in isolated compartments whose drains are designed to catch all potential leakage.
These drains are routed through the radioactive drainage systems to the waste management systems. Pumps whose potential for radioactive leakage is greatest are equipped with drip pans and lines piped to the floor drains. Discussion of the Equipment Drain System can be found in FSAR Subsection 9.3.3.
4) Cubicles where the potential for liquid leakage exists are generally provided with floor drains and/or equipment drains. Areas where flooding could be expected to cause a safety problem are provided with watertight doors.
The Waterford 3 Leak Reduction procedure addresses: (1) performance of inspections to verify integrity of systems that could cause an inadvertent release, and (2) implementation of a preventive maintenance program to promptly repair identified problems, such as leaking equipment and plugged floor drains.
Underground piping will be pressure tested as required by ASME Section XI or other regulatory requirements. New permanent piping systems will be pressure tested prior to first use in accordance with ASME.
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Section XI or other applicable regulatory requirements. All temporary piping associated with vendor solidification equipment is hydrostatically tested by the vendor prior to shipment and installation in accordance with their QA procedures.
1.9.38 IMPROVED IMPLANT IODINE INSTRUMENTATION UNDER ACCIDENT CONDITIONS (III.D.3.3)
Position
(1) Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during an accident.
(2) Each applicant for a fuel-loading license to be issued prior-to January 1, 1981 shall provide the equipment, training, and procedures necessary to accurately determine the presence of airborne radioiodine in areas within the plant where plant personnel may be present during an accident.
Response
Surveillance of inplant airborne Iodine concentration for the Waterford 3 SES is provided for via three different methods. The specific methods are:
1. In Plant Airborne Radiation Monitoring System (PARMS) (DRN 01-758) 2. Portable Continuous Air Monitors (PCAMS) (DRN 01-758) 3. Portable (hand-held) High Volume Air Samplers used in conjunction with the existing GE (Li) detector system which is located in the counting room
Method #1 is applicable for use only in the RAB whereas items 2 and 3 can be used by Health Physics Personnel anywhere in the plant where access may be required. An approximate indication of Fuel Handling Building airborne activities is provided for by two FHB normal exhaust monitors and four FHB isolation monitors described in FSAR Subsection 12.3.4.2.1. (DRN 02-406) Initial indication of potentially high airborne iodine concentration is provided for via stationary In Plant airborne Particulate Iodine and Gas Monitors. Four of these stationary monitors draw isokinetic samples of air from RAB ductwork. The sample points for three of the four RAB monitors were picked such that common exhaust ducts are sampled which are collecting exhausts from various rooms in the RAB in which occupancy is periodic as defined in FSAR Section 12.3A. (DRN 02-406) The fourth RAB In Plant Airborne Radiation Monitoring (IPSRM) draws a sample from the exhaust plenum of the RAB ventilation system and provides for overall monitoring of Particulate Iodine and Gas airborne concentrations in the RAB. (DRN 00-1053; 02-406) The remaining stationary In Plant Airborne Particulate Iodine and gas monitors provide for room specific particulate iodine and gas concentration readings. The specific areas are in the Hot Machine shop and Decontamination Area. (DRN 00-1053; 02-406)
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The iodine detector portion for all of the stationary monitors consists of a charcoal cartridge assembly and lead plug on the front end of the shield and a NaI (Tl) detector assembly and lead plug on the opposite end. The sample enters the shield, passes through the replaceable charcoal cartridge, and then exits the shield. The charcoal cartridge absorbs iodine and can be manually changed (in less than one minute). In operation, the cartridge is viewed with the NaI (Tl) integral line gamma scintillation detector. A single- channel analyzer (SCA) in the signal processor monitors an adjustable (set at 10 percent) window around the 364.KeV I131 photo peak.
During accident conditions Noble gases are also adsorbed on the charcoal and can "swamp" the iodine detector due to the poor resolution of NaI and because no background subtraction is provided, use of the above detectors results in overly conservative estimates of the airborne iodine concentrations when large amounts of noble gases are present. In order to minimize unnecessary usage of respirators, by plant personnel and in order to assist Health Physics personnel in quickly localizing the area from which high airborne, iodine concentration is arising, additional portable hand held high volume air samplers utilizing silver zeolite filters are available for use under accident conditions. Silver zeolite will retain iodine as well as, if not better than, charcoal but has low retention of Noble gases.
After securing a sample from the area of interest through the use of the high volume samplers Health Physics personnel would then take the sampler to the counting room where the Ge (Li) spectroscopy system is located for detailed spectroscopic analysis. A clean air source will be available for purging the silver zeolite cartridges of noble gases prior to counting. Appropriate precautions will be taken to prevent the spread of contamination during purging. As stated in FSAR Subsection 12.3A.3.3 post accident sampling shall not be performed from the existing sampling panel. As a result, the counting room background activity will be 2.6 mr/m 20 minutes after the accident and less than 1 mr/hr after six hours. At this level of background activity the existing Ge (Li) spectroscopy system is usable for accurately analyzing the iodine content of the sample filter. Utilizing the information obtained from the spectroscopic analysis of the sampler filter the Health Physics personnel at the Waterford 3 SES can accurately determine whether the use of respirators is required by plant personnel who would be entering the area from which the sample was taken.
The gamma-spectrometer consists of a computer based multichannel analyzer (MCA) which is used in conjunction with Ge (Li) detectors and with which it is fully compatible, and also consists of a preamplifier/amplifier for signal conditioning. A computer is utilized for data processing and storage.
The Ge (Li) detectors are the closed-end-coaxial type used for gamma spectroscopy for radiochemistry and health physics, and are mounted inside a chamber shielded with six inches of lead to minimize the effects of background on sample analysis. Software programs are available in the spectroscopy system which can assist the Health Physics Personnel in accurately determining iodine concentrations in the sample filters. Specifically the programs are: spectrum smoothing, peak search, and nuclide identification.
The Ge (Li) detectors have the following characteristics which assure accurate readings: a) Relative photopeak efficiency for 1.33 MeV photons: 15 percent for one detector and - 20 percent for the second detector.
1.9-41 Revision 14 (12/05)
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b) Peak to compton ratio - 40:1 c) Resolution (@ FWHM of Co60 1.33 peak MeV) £ 2.0 keV d) Full width at one tenth maximum (FWTM), is less than or equal to double the full width at half maximum (FWHM), i.e., FWTM £ 2 FWHM.
Once occupancy constraints have been established by Health Physics personnel through the use of portable high volume air samplers and the GE (Li) Gamma Scintillation detectors, Portable continuous airborne activity monitors would be utilized by personnel occupying the area. The function of the aforementioned Portable Continuous Air Monitors (PCAM) would be to monitor habitability conditions of the occupied area.
If high concentrations of airborne activities are detected the PCAMs shall alarm in the area. Personnel on hearing the alarm will evacuate the area. Health Physicists will then have to reestablish occupancy constraints through the use of the portable high-volume air samplers and the Ge(Li) spectroscopy system.
In summary monitoring of overall RAB radioactive particulate Iodine and gas concentratures is provided for via the IPARMS. Using these monitors the presence of potentially unacceptability high airborne activity levels can be detected in the RAB.
If the existence of a potential problem area is shown by the IPARMS, portable high volume samplers shall be used by Health Physics personnel to establish initial occupancy constraints for personnel who shall be working in a given area. Analysis of the samples taken via the portable high volume samplers shall be performed by the existing Ge (Li) spectoscopy system located in the counting room.
Finally, if plant personnel shall have to work, in an area that was previously cleared by Health Physics personnel for occupancy, for any length of time the PCAM's shall be placed into the work area to continuously monitor airborne activity levels. If unusually high activity levels were to be detected, the personnel occupying the area would be warned by the PCAM alarms.
Portable high volume air samplers and filter cartridges, and clean air purge capability meet the requirements of NUREG-0737.
The presence of airborne noble gases in the vicinity of the gamma spectroscopy equipment should not normally interfere with its ability to analyze the iodine content of the sample filter. However, an alternate gamma spectroscopy system will be available onsite and will be located in a more habitable counting room.
1.9-42 Revision 14 (12/05)
WSES-FSAR-UNIT-3
1.9.39 CONTROL-ROOM HABITABILITY REQUIREMENTS (III.D.3.4)
Position
Licensees shall assure that control room operators will be adequately protected against the effects of accidental release of toxic and radioactive gases and that the nuclear power plant can be safety operated or shut down under design basis accident conditions (Criterion 19, "Control Room," of Appendix A, "General Design Criteria for Nuclear Power Plants," to 10CFR Part 50).
Response
Waterford 3 has met the requirements of SRP 6.4, 2.2.1, 2.2.2, and 2.2.3 in previous submittals to the NRC. A control room habitability evaluation was performed in accordance with the requirements and clarifications of this position, and FSAR Table 1.9-2 identifies specific information requested or references FSAR sections where information has previously been submitted.
(DRN 04-1619, R14) The Waterford 3 control room is designed to be habitable during both toxic gas and radiological emergencies. The control room design meets the guidance of Regulatory Guide 1.78 (June 1974) and 1.95 (February 1975) as well as complies with GDC19 of Appendix A to 10CFR50 and 10CFR50.67. Control room habitability systems and equipment is discussed in Section 6.4 of the FSAR, while the toxic gas analysis is contained in Subsection 2.2.3. (DRN 04-1619, R14)
1.9-43 Revision 14 (12/05)
WSES-FSAR-UNIT-3 TABLE 1.9-1 (Sheet 1 of 5) Revision 308 (11/14) TMI-RELATED REQUIREMETNS FOR NEW OPERATING LICENSES No. Title Description FSAR Sections References I.A.1.1 Shift Technical Advisor 1. On Shift Subsection 13.1.2.2.2 (1) item 2.2.lb, Responsibilities 2. Training (2), (3), (4) 3. Describe long term program (LBDCR 13-015, R308) I.A.1.2 Shift Manager Delegate Non-Safety Subsection 1.9.2 (1) item 2.2.1a, Responsibilities Duties (2), (3) (5) (LBDCR 13-015, R308) I.A.1.3 Shift Manning 1. Limit overtime 1. Subsection 1.9.3 2. Minimum Shift Crew 2. Table 16.6.2-1, & Table 6.2-1 I.A.2.1 Immed. Upgrade of RO & SRO Training 1. SRO Experience Subsections 13.2.1, 13.2.2 (4), (6), (7) and Qualifications 2. SROs be ROs 1 year 3. Three Mos. tng. on-shift 4. Modify Training 5. Facility Certification I.A.2.3 Administration of Training Programs Instructors complete Subsections 13.2.1, 13.2.2, (4), (6) SRO exam 13.2.4, 13.3.7 I.A.3.1 Revise Scope and Criteria for 1. Increase Scope Subsection 13.2.2 (4), (6), (7) Lic. Exams 2. Increase Passing Grade 3. Simulator Exams a) Plants with Simulators b) All Plants I.B.1.2 Evaluation of Organization Organization, resources Subsection 1.9-7, (7) tng. and qualifications Chapter 13 for operators and accidents I.C.1 Short-term Accident 1. SB LOCA Subsections 1.9.8, 6.3 (1) item 2.1.3b and Procedure Review 2. Inadequate Core Cooling 13.2.1, 15.6.3 2.1.9, (2) a) Reanalyze and Propose (3), (4) Guidelines b) Revise Procedures 3. Transients & Accidents a) Reanalyze and Propose Guidelines b) Revise Procedures I.C.2 Shift and Relief Revise procedures to assure Subsection 1.9.9 (1) item 2.2.1c, Turnover Procedures plant status known by new shift (2), (3) I.C.3 Shift Supervisor Corporate directive to establish Subsection 1.9-10 (1) item 2.2.1a, command duties and revise plant (2), (3) procedures WSES-FSAR-UNIT-3 TABLE 1.9-1 (Sheet 2 of 5) TMI-RELATED REQUIREMETNS FOR NEW OPERATING LICENSES No. Title Description FSAR Sections References I.C.4 Control Room Access Establish authority Subsection 1.9.11 (1) item 2.2.2a, and limit access (2), (3) I.C.5 Feedback of Operating Experience Review and Revise Procedures Subsection 1.9.12 (4),(7) I.C.6 Verify Correct Performance of Revise Performance Procedures Subsection 1.9.13 (4) Oper. Activites I.C.7 NSSS Vendor Review of Procedures 1. Low Power Test Program Subsections 1.9.14, (4),(7) 2. Power Ascension Procedures 13.5.1.2, 14.2.2.5, 14.2.3 3. Emergency Procedures I.D.1 Control Room Design Rev. Preliminary Assessment and Subsection 1.9.15 (7) schedule for correcting deficiencies I.D.2 Plant-Safety- Parameter Display 1. Description Subsection 1.9.16 Console 2. Installed 3. Fully Implemented I.G.1 Training During Low 1. Propose Tests Section 14.2, & Power Testing 2. Submit Analysis and Procedures Subsection 13.2.1.2 3. Training Results II.B.1 Reactor Coolant System Vents 1. Design Analysis Subsection 5.4.15 (2), (3), (4) 2. Install 3. Procedures II.B.2 Plant Sheilding 1. Radiation and Shielding Review Appendix 12.3A (1) item 2.1.6.6, 2. Corrective Actions to Subsection 9.3.8 (2), (3), (4) Assure Access 3. Complete Mods 4. Equipment Qualifications II.B.3 Post Accident Sampling 1. Design Review (1) item 2.1.8a, 2. Corrective Actions (2),(3),(4) 3. Procedures 4. Complete Actions II.B.4 Training for Mitigating 1. Develop Training Program Subsections 13.2.1, 13.2.2 (6),(7) Core Damage 2. Complete Training II.D.1 Relief and Safety Valve Test Requirements 1. Describe Program and 1. Subsection 1.9.22 (1) item 2.1.2, Schedule WSES-FSAR-UNIT-3 TABLE 1.9-1 (Sheet 3 of 5) TMI-RELATED REQUIREMETNS FOR NEW OPERATING LICENSES No. Title Description FSAR Sections References II.D.1 (Cont’d) 2. RV and SV Tests 2. Subsection 1.9-22 3. Block Valve Tests 3. (NA) II.D.3 Valve Position Indication Install In Control Room 1.9.23 (1) item 2.1.3a (2),(3),(4)
II.E.1.1 AFV System Evaluation 1. Analysis Appendix 10.4.9 A & 9B (8) 2. Modification II.E.1.2 AFW System Initiation and Flow 1. Initiation Subsection 7.3.1.1.6, (1) item 2.1.7a & (a) Control grade Table 7.5-1, (2),(3) (b) Safety grade 10.4.9, Table 1.7-1 2. Flow Indication (a) Control grade (b) Safety grade II.E.3.1 Emergency Power for Pressurizer Heater Installed Capability Subsections 1.9.26 and (1) item 2.1.1 5.4.1.0.2, Figure 8.3-33 (2),(3),(4) II.E.4.1 Dedicated Hydrogen Penetrations 1. Design 1. (NA) (1) item 2.1.5a 2. Review and Revise 2. Subsections 1.9.27, 6.2.5 2.1.5c, H2 Control Proc 3. (NA) (2),(3),(4) 3. Install II.E.4.2 Containment Isolation Dependability 1-4.Implement diverse isolation 1) Subsection 1.9.28 (1) item 2.1.4, 5. Cont. pressure setpoint Table 1.9-3, (2),(3),(4) 6. Cont. purge valve Table 6.2-32 7. Radiation signal on purge valve II.F.1 Accident Monitoring Instrumentation 1. Procedures 1. Subsection 1.9.2.9 (1) item 2.1.8b, 2. Install Instrumentation 2. (a), (b), (c), (d), (e) (2),(3),(4) (a) Noble gas monitor Subsection 1.9.29 (b) Iodine/particulate sampling (c) Containment high range (f) Subsection 6.2.5.1 monitor (d) Containment pressure (e) Containment water level (f) Containment hydrogen WSES-FSAR-UNIT-3 TABLE 1.9-1 (Sheet 4 of 5) TMI-RELATED REQUIREMETNS FOR NEW OPERATING LICENSES No. Title Description FSAR Sections References II.F.2 Instrumentation for 1. Procedures Subsection 1.9.30 (1) item 2.1.sb, detection of inadequate 2. Subcooling meter (2),(3),(4) core-cooling 3. Describe other instrumentation 4. Install add’l instrumentation II.G.1 Power supplies for pressurizer relief valves, Power supply from emergency buses Table 1.7-1 (1) item 2.1.1, Subsections 1.9.31 (2), (3), (4) 8.3.1.1.1c, 7.7.1.2.2 II.K.1 IE Bulletins 5. Review ESF Valves 5. Subsection 1.9.32 (7) 10. Operability Status 10. Subsection 1.9.32 II.K.2 Orders on B & W Plants 13. Thermal-mechanical Report 13. Subsection 1.9.33 17. Voiding in RCS 17. Subsection 1.9.33 19. Benchmark analysis 19. Subsection 1.9.33 seq AFW flow II.K.3 Final Recomendations, B & O Task Force 1. Auto PORV isolation 1. (NA) (4),(7),(8) 2. Report PORV failures 2. (NA) 3. Reporting SV and RV 3. Subsection 1.9.34 failures & challenges 5. Auto trip of RCPs 5. Subsection 1.9.34 a) propose mods b) modify 7. Evaluation of PORV opening probability 17. Subsection 1.9.34 17. ECCS outages 25. Subsection 1.9.34 25. Power on pump seals a) propose mods b) mods II.K.3 Cont’d 30. SB LOCA Methods 30. Subsection 1.9.34 a) schedule outline b) model c) new analysis 31. Plant Specific Analysis 31. Subsection 1.9.34 III.A.1.1 Emergency Preparedness, 1. Comply with 1OCFR5O, APP. E Subsection 1.9.35 and (7) Short term 2. Comply with NUREG-0654 Section 13.3 3. Conduct Exercise 4. Meteorotogical Data WSES-FSAR-UNIT-3 TABLE 1.9-1 (Sheet 5 of 5) TMI-RELATED REQUIREMETNS FOR NEW OPERATING LICENSES No. Title Description FSAR Sections References III.A.1.2 Upgrade Emergency 1. Establish (Interim Basis) Subsections 1.9.36, (1) item 2.2.2b, Support Facilities (a) TSC 13.3.6 2.2.c, (2), (3) (b) OSC (c) EOF 2. Design 3. Modifications III.A.2 Emergency Preparedness Long Term 1. Upgrade Emergency Plan Subsection 1.9.36a & Long Term to APP. E, 10CFR50 Section 13.3 2. Meterorological Data III.D.1.1 Primary Coolant Measure leak rates and establish Subsections 1.9.37, (1) item 2.1.6a, Outside Containment program to keep leakage ALARA 6.2.2.4.2, 6.3.4.3, (2),(3),(4) 9.3.4.3.4, 9.3.6.4 III.D.3.3 Inplant 1, Radiation 1. Provide means to Subsection 1.9.38 (1) item 2.1.8c, Monitoring determine presence (2),(3),(4) of radio-iodine 2. Modifications to accurately measure radio-iodine III.D.3.4 Control Room 1. Identify and evaluate Subsection 1.9.39, Habitability potential hazards Table 1.9-2 2. Schedule for Modifications 3. Modifications
WSES-FSAR-UNIT-3
References To Table 1.9-1
1. Nuclear Regulatory Commission, "TMI-2 Lessons Learned Task Force Status Report and Short- Term Recommendations," USNRC Report NUREG-0578, July 1979.
2. Letter from D. B. Vassallo, NRC, to All Pending Operating License Applicants, Subject: Followup Actions Resulting from the NRC Staff Reviews Regarding the Three Mile Island Unit 2 Accident, dated September 27, 1979.
3. Letter from D. B. Vassallo, NRC, to All Pending Operating License Applicants, Subject: Discussion of Lessons Learned Short-Term Requirements, dated November 9, 1979.
4. Letter from D. G. Eisenhut, NRC, to All Licensees of Operating Plants and Applicants for Operating Licenses and Holders of Construction Permits, Subject: Preliminary clarification of TMI Action Plan Requirements, dated September 5, 1980.
5. Letter from D. G. Eisenhut, NRC, to All Licensees of Operating Plants and Applicants for Operating Licenses and Holders of Construction Permits, Subject: Interim Criteria for Shift Staffing, dated July 31, 1980.
6. Letter from H. R. Denton, NRC, to All Power Reactor Applicants and Licensees, Subject: Qualification of Reactor Operators, dated March 28, 1980.
7. Nuclear Regulatory Commission, "TMI-Related Requirements for New Operating Licenses", USNRC Report NUREG-0694,June 1980.
8. Letters from D. F Ross, Jr., NRC, to All Pending W, CE, and B&W License Applicants, Subject: Actions Required from B&O Task Force Review, dated March 10, 1980 (April 24, 1980).
WSES FSAR UNIT 3
TABLE 1.9 2 (Sheet 1 of 4) Revision 301 (09/07)
TMI INFORMATION REQUIRED FOR CONTROL ROOM
HABITABILITY EVALUATION (TASK ACTION PLAN ITEM III.D.3.4)
ITEM NO. Information Required Information Reference 1. Control room mode of operation i.e. pressurization See reference Subsection 6.4.3.3 and filter recirculation for radiological accident isolation or chlorine release.
G(EC 5000082445, R301) 2. Control room characteristics 220,000 ft3 Maximum Subsection 6.4.2.2 a) air volume control room 168,500 ft3 Minimum and Table 6.4 2 0(EC 5000082445, R301)
b) Control room emergency zone (control room, The control room envelope is defined to Subsection 6.4.2.1 critical files and kitchen, washroom, computer include the main control room, computer room, and figure 1.2 8 computer room air conditioning equipment room, etc). room, control room HVAC equipment room, emergency living quarters, emergency food and water storage rooms, toilets, locker room, kitchenette, supervisors office, corridors, conference room and vault (critical document reference file).
Control room operators will require access to the above areas immediately after and during an emergency.
The entire envelope floor is at elevation +46 ft. MSL inside the Reactor Auxiliary Building.
Drawing G134 is a layout drawing showing the control room envelope, and the placement of equipment.
G(DRN 07 241, R301) c) Control room ventilation system schematics with See reference Figure 6.4 1 through normal and emergency air flowrates. 6.4 3 0(DRN 07 241, R301) WSES FSAR UNIT 3
TABLE 1.9 2 (Sheet 2 of 4) Revision 11 A (02/02)
TMI INFORMATION REQUIRED FOR CONTROL ROOM
HABITABILITY EVALUATION (TASK ACTION PLAN ITEM III.D.3.4)
ITEM NO. Information Required Information Reference d) Infiltration leakage rate 200CFM (max) Subsection 6.4.2.3 and Table 6.4.2
¨ (DRN 01 570) e) high efficiency particulate air (HEPA) filter and HEPA filter 99.97% Table 9.4 2 charcoal absorber efficiencies Charcol Absorber see referenced table õ (DRN 01 570) f) Closest distance between containment and air Approximately 75 ft. Figure 1.2 1 intake
g) Layout of control room, air intakes, containment See reference Subsection 6.4.4.2, building, and chlorine, or other chemical storage Table 2.2 3, 1.1 2, 1.2 facility with dimensions. 8, 1.2 17, 1.2 18, 1.2 19, 1.2 20, 1.2 21 and 1.2 22
h) control room shielding including radiation See reference Subsection 6.4.2.5 streaming from penetrations, doors, ducts, stairways, etc.
i) Automatic isolation capability damper closing See reference Subsection 6.4.4.2 time, damper leakage and area. and Table 6.4 1
j) Chlorine detectors or toxic gas (local or remote) Chlorine and Anhydrous Ammonia detectors are Subsection 6.4.4.2 local detectors
k) self contained breathing apparatus availability 7 units (for 5 men) Subsection 6.4.4.2 (number)
l) bottled air supply (hours supply) 6 hours (for 5 men) Subsection 6.4.4.2 WSES FSAR UNIT 3
TABLE 1.9 2 (Sheet 3 of 4) Revision 11 (05/01)
TMI INFORMATION REQUIRED FOR CONTROL ROOM
HABITABILITY EVALUATION (TASK ACTION PLAN ITEM III.D.3.4) ITEM NO. Information Required Information Reference m) Emergency food and potable water supply (how 5 days (for 5 men) Subsection 6.4.4.2 many days and how many people)
n) Control room personnel capacity (normal and 5 men Subsections 6.4.1 and emergency) 6.4.4.2
o) potassium iodide drug supply See reference Subsection 13.3.5.6.2
ç (DRN 99 1093) Ï (DRN 99 1093) 4. Offsite manufacturing, storage or transportation See reference Subsection 2.2.2, facilities of hazardous chemicals. 2.2.3 and Tables 2.2 1 through 2.2 10
a) identify facilities within a five mile radius See reference Table 2.2 1
b) Distance from control room See reference Table 2.2 3
c) Quantity of hazardous chemicals is one container See reference Table 2.2 3
d) Frequency of hazardous chemical transportation See reference Subsections 2.2.2.1.2 traffic (truck, rail, and barge) 2.2.2.2, 2.2.2.3, 2.2.2.4, Tables 2.2 2, 2.2 4, 2.2 5, 2.2 6, 2.2 8 and 2.2 9 WSES FSAR UNIT 3
TABLE 1.9 2 (Sheet 4 of 4) Revision 11 (05/01)
TMI INFORMATION REQUIRED FOR CONTROL ROOM
HABITABILITY EVALUATION (TASK ACTION PLAN ITEM III.D.3.4)
ITEM NO. Information Required Information Reference 5. Technical Specifications (refer to standard technical specifications)
a) Chlorine detection system For chlorine or ammonia detectors, see Subsection reference 16.2.3/4.3.3.7
b) control room emergency filtration system See reference Subsection including the capability to maintain the control 16.2.3/4.7.7.1 room pressurization at 1/8 in. water gauge, verification of isolation by test signals and damper closure times, and filter testing requirements.
WSES-FSAR-UNIT-3
TABLE 1.9-3
CONTAINMENT ISOLATION VALVES PROVIDED WITH CAPABILITY FOR MANUAL OVERRIDE
System Name Fluid Valve Tag No.
1. Instrument Air Compressed Air 2IA-F601A/B
2. Component Cooling Demineralized Water 2CC-F146A/B Water Inlet to Reactor Coolant Pumps and CEDM Coolers
3. Component Cooling Demineralized Water 2CC-F147A/B Water Outlet from 2CC-243A/B Reactor Coolant Pumps and CEDM Coolers
4. CVCS Letdown Borated Water 2CH-F1518A/B Line 1CH-F2501A/B
5. Sampling Line from Primary Coolant 2SL-F1504A/B RCS 2SL-F15O1A/B
6. Sampling Line from Primary Coolant 2SL-F1505A/B Pressurizer Surge 2SL-F15O2A/B
7. Sampling Line from Primary Coolant 2SL-F1506A/B Pressurizer Steam Space 2SL-F1503A/B
8. Containment Sump Pump Borated Water 2WM-F105A/B Discharge 2WM-F104A/B
9. CVCS from RCP Controlled Primary Coolant 2CH-F1512A/B Bleedoff 2CH-F1513A/B
10. Sampling from Steam Secondary Coolant 2SL-F602 Generator 1 Blowdown 2SL-F601
Sampling from Steam Secondary Coolant 2SL-F604 Generator 2 Blowdown 2SL-F603
12. Hydrogen Analyzer Hydrogen 2HA-E609A Supply and Return Lines 2HA-E608A 2HA-E610A 2HA-E629B 2HA-E628B 2HA-E630B
WSES-FSAR-UNIT-3
TABLE 1.9-4 Revision 10 (10/99)
ADDITIONAL ACCIDENT MONITORING INSTRUMENTATION
1 23456 78910 11 12 131415161718 Control Wiring General Qualified LOCATION - FSAR Fig. No./Column Line Calibra- Calibra-
Detector Dia. Atomic to IEEE Iso- Micro- Energy Response tion tion Radiation Monitoring Instru. B-424 Model No. Of Power 323/74 nozzle Filter Detector processor Readout Sensi- Depend- cpm/ Frequency Method- Instrumentation Tag No. Sh. No. No. Channels Source 344/75 Recording Range tivity ence ?Ci/cc (Note 1) ology ç
1. Plant Vent Stack RD-HV-0110 2648 RD-52 1 Note 3 Yes EL. + 111 1.2-17 1.2-17 Dwg. G135 CR Panel CR-CP-52 10 7-105 NA NA 4.32 x 107 1Note 2 RD-72 Plant Stack 5A 8A 7A-8A/ CP-52 & with μCi/cc 2.84 x 104 K-L CP-6 rec.(Note 4) 1.19 x 102 Ï 2. Condensor Vacuum RD-AE-0002 2649 RD-52 1 Note 3 Yes NA 1.2-3 1.2-3 1.2-3 CR Panel CR-CP-52 10 7-105 NA NA 4.32 x 107 1Note 2 Pump Effluents RD-72 7G 6G 5G Cp-51 with μCi/cc 2.84 x 104 CP-52 & rec.(Note 4) 1.19 x 102 CP-6
3. FHB Emergency RD-HV- 2639 RD-52 1 Note 3 Yes FHB EL. 1.2-15 1.2-17 1.2-15 CR Panel CR-CP-52 10 7-105 NA NA 4.32 x 107 1Note 2 Exhaust 3032 RD-72 9 ft. 6 3FH/V-T V CP-51 with μCi/cc 2.84 x 104 in.Exh. CP-52 & rec.(Note 4) 1.19 x 102 Plenum CP-6 ç 4. Main Steam Line RD-MS- 2690 2 Note 4 Yes NA NA 1.2-17 Dwg. G135 CR Panel CP- CR-CP-52 NA 100KeV NA 1 Note 2 5500 A, 2691 (1 for 140° 7A-8A/ 51 with -3MeV 5500 B each M.S. K-L CP-52& rec. ± 20% line) CP-6
5. Containment RD-CA-5400 2635 S RS 23A 2 Note 5 Yes NA NA 1.2-17 Dwg. G135 CR Panel CR-CP-14 100-108 NA 0.1-3MeV NA 1 Note 2 High Range AS 2636 S (SA, SB) 270°- 7A-8A/ CP-14 & with mr/hr ± 20% 5400 BS 90° K-L CP-6 rec. Ï
NOTES 1. Calibration Frequency per refueling outage as recommended by NUREG 0472. 2. Methodology submitted by manufacturer - to be incorporated into Operating Procedures. 3. Power not available upon Loss of Offsite Power (LOOP). 4. Power from Static Uninterruptible Power Supply (SUPS). 5. Power from Emergency DG upon LOOP.
WSES-FSAR-UNIT-3
APPENDIX 1.9A
RESPONSE TO SECTION II.F.2 OF NUREG-0737 INADEQUATE CORE COOLING INSTRUMENTATION
TABLE OF CONTENTS
Section Title Page 1.9A.1 INTRODUCTION 1.9A-1 1.9A.1.1 DEFINITION OF ICC 1.9A-1 1.9A.1.2 DESCRIPTION OF ICC EVENT PROGRESSION 1.9A-1 1.9A.1.3 SUMMARY OF SENSOR EVALUATIONS 1.9A-2 1.9A.2 FUNCTIONAL DESCRIPTION OF ICCI 1.9A-2 1.9A.2.1 INTERVAL 1 - APPROACH TO SATURATION 1.9A-2 1.9A.2.2 INTERVAL 2 - APPROACH TO CORE UNCOVERY 1.9A-2 1.9A.2.3 INTERVAL 3 - CORE UNCOVERY 1.9A-3 1.9A.2.4 INTERVAL 4 - RECOVERY FROM ICC 1.9A-3 1.9A.3 ICCI SENSOR DESIGN DESCRIPTION 1.9A-3 1.9A.3.1 SATURATION MARGIN MONITOR 1.9A-4 1.9A.3.2 HEATED JUNCTION THERMOCOUPLE SYSTEM 1.9A-4 1.9A.3.3 CORE EXIT THERMOCOUPLE SYSTEM 1.9A-6 1.9A.4 SIGNAL PROCESSING AND DISPLAY 1.9A-7 1.9A.4.1 QSPDS PROCESSING 1.9A-8 1.9A.4.2 QSPDS DISPLAY 1.9A-12 1.9A.5 SYSTEM VERIFICATION TESTING 1.9A-12 1.9A.5.1 RTD AND PRESSURIZER PRESSURE SENSORS 1.9A-12 1.9A.5.2 HJTC SYSTEM 1.9A-12 1.9A.5.3 CETs 1.9A-13 1.9A.6 ICCI SYSTEM QUALIFICATION 1.9A-13 1.9A.7 OPERATING INSTRUCTIONS 1.9A-13
1.9A-i
WSES-FSAR-UNIT-3
APPENDIX 1.9A
TABLE OF CONTENTS (Cont'd)
Section Title Page
1.9A.8 COMPARISON OF DOCUMENTATION REQUIREMENTS WITH THIS REPORT 1.9A-14 1.9A.9 SCHEDULE FOR ICCI IMPLEMENTATION 1.9A-14 (DRN 01-758) 1.9A REFERENCES TO APPENDIX 1.9A 1.9A-15 (DRN 01-758)
1.9A-ii Revision 11-A (02/02)
WSES-FSAR-UNIT-3
APPENDIX 1.9A
RESPONSE TO SECTION II.F.2 OF NUREG -0737 INADEQUATE CORE COOLING INSTRUMENTATION
LIST OF TABLES
Table Title 1.9A-1 Definition of ICC Event Progression Intervals 1.9A-2 Comparison of ICCI to Documentation Requirements of Item II.F.2 of NUREG-0737 1.9A-3 Comparison of ICCI to Attachment 1 of II.F.2 1.9A-4 Comparison of ICCI to Appendix B of NUREG-0737
1.9A-iii Revision 11-A (02/02)
WSES-FSAR-UNIT-3
APPENDIX 1.9A
RESPONSE TO SECTION II.F.2 OF NUREG -0737 INADEQUATE CORE COOLING INSTRUMENTATION
LIST OF FIGURES
Figure Title 1.9A-1 Definition of Intervals in Event Progression 1.9A-2 ICC Instrumentation System 1.9A-3 HJTC Sensor 1.9A-4 HJTC Split Probe Design Configuration 1.9A-5 HJTC Sensor Axial Locations 1.9A-6 Inadequate Core Cooling Reactor Vessel Level Cable Routing 1.9A-7 Core Exit Temperature Measurement Scheme 1.9A-8 Core Exit Thermocouple Core Locations 1.9A-9 Inadequate Core Cooling Core Exit Thermocouples Cable Routing 1.9A-10 HJTC Level Logic 1.9A-11 Heater Power Control Logic
1.9A-iv Revision 11-A (02/02)
WSES-FSAR-UNIT-3
APPENDIX 1.9A
RESPONSE TO SECTION II.F.2 OF NUREG -0737 INADEQUATE CORE COOLING INSTRUMENTATION
1.9A.1 INTRODUCTION
This document provides the response to the requirements of Section II.F.2 of NUREG-0737 (Reference 1) related to inadequate core cooling instrumentation (ICCI). Included is a description of the activities conducted by LP&L and the CE Owners Group (CEOG) to define and implement an ICCI System for Waterford 3. The report describes the instrumentation package selected by LP&L to provide an indication of the approach to, the existence of, and the recovery from inadequate core cooling (ICC). The ICCI System design is based on typical accident events which progress toward the defined state of ICC.
1.9A.1.1 DEFINITION OF ICC
(DRN 03-2054, R14) The criteria for the existence of ICC is based on the potential for significant core damage and fission product release to occur. ICC is defined to exist if the fuel clad temperature reaches or exceeds 2200F. This is the licensing clad temperature limit for design basis events analyzed in the FSAR. ICC can occur only if there is a significant loss of water inventory from the Reactor Coolant System (RCS) so that the coolant level drops below the top of the core. (DRN 03-2054, R14)
1.9A.1.2 DESCRIPTION OF ICC EVENT PROGRESSION
The evaluation of the instrument sensors to determine ICC is based on events which proceed slowly enough for the operator to observe and make use of the instrument displays. A small break LOCA illustrates the progression of such an event which can lead to ICC. Figure 1.9A-1 shows the representative behavior of the two-phase mixture level, RCS pressure, steam and clad temperatures with time for this event. The event progression is divided into four intervals shown in the figure. Any event which leads to ICC progresses through these intervals. The ICCI package is designed to provide information about each interval and therefore covers the entire event progression. (DRN 01-758, R11-A) The intervals of an ICC event progression are described in Table 1.9A-1. Interval 1 is characterized by a reduction in RCS subcooling until saturation occurs. This can occur by depressurization or by increasing the temperature of the RCS. During Interval 2, the coolant level in the reactor vessel falls to the top of the core as a result of a loss of coolant inventory. In Interval 3 the two-phase mixture level drops below the top of the core to its lowest level during the event progression. During this time the clad temperature increases and produces an increasing superheated steam temperature at the core exit. The final Interval 4 is the recovery from ICC. The two-phase level increases above the top of the core causing the clad and core exit steam temperatures to decrease. Complete recovery occurs when the reactor vessel is filled again (depending on break size) and RCS subcooling is established. (DRN 01-758, R11-A)
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1.9A.1.3 SUMMARY OF SENSOR EVALUATIONS
Several sensors have been evaluated for use in an ICCI System. The results of this evaluation are presented in CEN-117 (Reference 2). Based on this and other studies performed by the CEOG, LP&L has selected the following sensors for an ICCI package:
1) Hot and Cold Leg Resistance Temperature Detectors (RTDs)
2) Pressurizer Pressure
3) Heated Junction Thermocouples (HJTCs)
4) Unheated Junction Thermocouples (UHJTCs)
5) Core Exit Thermocouples (CETs)
1.9A.2 FUNCTIONAL DESCRIPTION OF ICCI
This section gives a functional description of the parameters which, when measured and displayed, provide the operator with advanced warning of the approach to, existence of and recovery from ICC. The key parameters for each interval of the ICC event progression are identified.
1.9A.2.1 INTERVAL 1 - APPROACH TO SATURATION
As described before, Interval 1 is the loss of RCS subcooling until saturation conditions are reached. The parameters measured to detect subcooling are the RCS coolant temperature and pressurizer pressure. With this information, the amount of subcooling and the occurrence of saturation conditions can be determined. Temperature is measured in the hot and cold legs, at the core exit, and in the reactor vessel upper head region. The measurement range extends so that saturation can be determined from shutdown cooling conditions up to the pressurizer safety valve setpoint pressure. The response time is adequate for the operator to obtain information during those events which proceed slowly enough for him to observe and take actions based on the indication.
1.9A.2.2 INTERVAL 2 - APPROACH TO CORE UNCOVERY
During Interval 2, the RCS remains at saturation conditions as coolant inventory is lost and the coolant level in the reactor vessel decreases. In order to track the continued progression of the event, an indication of the loss of inventory (liquid mass) prior to core uncovery is required. This is achieved by measuring the collapsed liquid level in the reactor vessel above the fuel alignment plate. The collapsed level is the level that results when all the voids (steam bubbles) in a two-phase mixture are collapsed. Measurement of the collapsed level, rather than the two-phase level, is more desirable since it provides a direct indication of the amount of liquid mass that exists in the reactor vessel above the core.
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The collapsed level is measured over the range of fluid conditions from shutdown cooling to saturation at the pressurizer safety valve setpoint pressure. The level measurement range extends from the top of the reactor vessel to the top of the fuel alignment plate. The response time is short enough to track the level during a small break LOCA. The measurement resolution is sufficient to indicate the progression of the event and the consequences of any mitigating action.
1.9A.2.3 INTERVAL 3 - CORE UNCOVERY
Interval 3 is characterized by an increasing fuel clad temperature caused by the two-phase mixture level falling below the top of the core. As the clad temperature increases, steam leaving the core becomes superheated. The amount and trend of the steam superheat provides an indication of the clad temperature and therefore, an indication of the approach to, or existence of ICC. Indication of the trend (increasing or decreasing) of the clad temperature is more important to the operator than information on the absolute value of the clad temperature since the trend tells him if conditions are getting better or worse.
(DRN 03-2054, R14) The core exit steam temperature is measured by thermocouples at an elevation just above the fuel alignment plate. The temperature range extends from saturation at shutdown cooling conditions to greater than the maximum predicted core average steam exit temperature which occurs when the peak clad temperature reaches 2200F. The range for processing of the thermocouple output extends to 2300F, although reduced accuracy is expected at the higher temperatures. (DRN 03-2054, R14)
1.9A.2.4 INTERVAL 4 - RECOVERY FROM ICC
Interval 4, recovery from the ICC event, begins after the two-phase mixture level in the core reaches a minimum and starts to increase. The increasing mixture level results in a decreasing core exit steam temperature until saturation temperature is reached when the core is completely covered. Measurement of the collapsed water level above the core provides continuous monitoring of the increasing inventory and recovery from ICC. Finally, subcooling of the RCS is re-established.
The parameters which indicate the recovery from ICC during Interval 4 are the same as those discussed for the first three intervals. Thus, the entire ICC event progression can be monitored by the operator.
1.9A.3 ICCI SENSOR DESIGN DESCRIPTION
The following instruments have been selected by LP&L to make up the ICCI System in response to NUREG-0737. This instrument package meets the functional requirements described in Section 1.9A.2.
1) Saturation Margin Monitor (SMM)
2) Heated Junction Thermocouple (HJTC) System
3) Core Exit Thermocouple (CET) System
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Figure 1.9A-2 shows a functional diagram for the ICCI System. Each instrument system consists of two safety grade channels from the sensors through the display.
1.9A.3.1 SATURATION MARGIN MONITOR
The SMM is a two channel, on-line system which provides a continuous indication of the RCS margin from saturation conditions (subcooled or superheated). It can be used to inform the operator of the approach to saturation and the existence of core uncovery. RCS pressure input to the SMM is provided by two (one per channel) wide range safety grade pressurizer pressure channels. RCS temperature inputs are provided by hot and cold leg RTD's, maximum unheated junction thermocouple (UHJTC) temperature from the upper head region, and the representative (maximum) core exit thermocouple temperature. The signals from the transmitters are carried to the QSPDS panels in two trains of Class 1E cabling, raceways and containment electrical penetrations. Each redundant train carries signals from two cold leg temperature, one hot leg temperature and one pressurizer pressure transmitters. The sensor inputs to the SMM are summarized below.
Input Range
Pressurizer Pressure 0-3000 psia
(DRN 03-2054, R14) Cold Leg Temperature (Ch. A-Loop, IA, 2A) (Ch. B-Loop, IB, 2B) 50-750F
Hot Leg Temperature (Ch. A-Loop 1) (Ch. B-Loop 2) 50-750F
Maximum UHJTC Temperature (from upper head) 32-2300F*
Representative CET Temperature 32-2300F* (DRN 03-2054, R14)
1.9A.3.2 HEATED JUNCTION THERMOCOUPLE SYSTEM
The principal function of the HJTC System is to measure the water inventory in the reactor vessel above the fuel alignment plate. This is done at discrete elevations by monitoring the temperature difference between adjacent heated and unheated thermocouples.
(DRN 03-2054, R14) ______* Thermocouples continue to function up to 2300F, although their accuracy is reduced above 1800F. (DRN 03-2054, R14
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The HJTC sensor, shown on Figure 1.9A-3 consists of two thermocouple junctions separated by several inches and a splash shield. One of the junctions is heated by an electric coil. When the heated junction is surrounded by a fluid of relatively good heat transfer properties (liquid), the temperature difference between the two thermocouple junctions is small (less than 2000F) When the heated junction is surrounded by a fluid of poor heat transfer properties (steam), the temperature difference is large (much greater than 2000F). Thus, by monitoring the temperature difference between adjacent heated and unheated thermocouples, it can be determined if an individual HJTC sensor is covered by liquid or surrounded by steam. The splash shield protects the heated junction from spurious cooling by water running down the sensor sheath or entrained water droplets.
Eight HJTC sensors are placed at specific elevations inside a separator tube to make up a probe assembly. The purpose of the separator tube is to create a collapsed water level inside while a two-phase mixture exists outside the tube. When the collapsed water level falls below a heated junction elevation, its temperature and the sensor differential temperature increase above a predetermined setpoint value. The sensor is then identified as being uncovered (i.e., surrounded by steam).
At Waterford 3, a "split probe" configuration is used. This refers to the separator tube which is divided into two independent separator tubes, one on top of the other, each of which creates a collapsed level inside it (see Figure 1.9A-4). A divider disk inside the separator tube located at the elevation of the upper guide structure support plate hydraulically isolates the two regions. Thus, the collapsed water level is measured in the upper plenum as well as, and separately from, the collapsed water level in the upper head. Each portion of the split probe has eight holes of 13/64 inch diameter near both the top and the bottom. This provides approximately the same flow area for water drainage as was used and verified to be adequate in the Phase 11 tests of the HJTC probe assembly (Reference 3).
Two independent HJTC split probe assemblies are installed in the reactor vessel. They are located near the periphery of the upper guide structure and away from the hot legs. Each probe assembly is housed within a stainless steel guide tube which protects it from hydraulic loads and serves as a guide path for the probe. A third tube, between the upper guide structure support plate and the fuel alignment plate, provides additional support and attaches the entire assembly to a control element assembly shroud. The guide and support tubes are perforated along their entire length with 3/8 inch holes. Additionally, slots in these tubes are positioned relative to the holes in the separator tube so as to prevent steam bubbles from entering the probe at the bottom and entrained water droplets from entering the top. The response to Question 10 in CEN-181 (Reference 4) provides more details on this arrangement.
The axial location of each HJTC sensor is shown on Figure 1.9A-5. The location of the sensors is identical for both of the instrument probe assemblies. Thus, failure of any one sensor does not decrease the measurement resolution since a sensor at the same elevation in the second probe provides the same information. Three sensors are located in the top head region and five in the upper plenum. In each region, sensors are placed as high and as low as possible to inform the operator when the region is completely full or empty. Sensors in between provide additional resolution and information on the progress of the collapsed level during an ICC event.
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The response time for the HJTC System is given in CEN-185, Supplement 3 (Reference 5). For a decreasing water level (drain), the response time is governed by the heat-up rate of the heated thermocouple which varies with pressure (and heater power). It is sufficiently short to inform the operator in a timely fashion that inventory has been lost. For an increasing water level (refill), the response time is much shorter since quenching the heated thermocouple removes heat quickly.
A sensor heater power control system is used to protect the heated junction thermocouple from damage due to overheating. When an increasing heated junction temperature or sensor differential temperature exceed a preset value the heater power is reduced until an acceptable stable temperature is reached. The power still remains high enough however, so that all sensors are capable of providing an uncovered signal. A more detailed description of the heater power control scheme is given in Reference 5.
The signal from each HJTC and heater circuits is carried from the reactor vessel head area to the QSPDS panels in redundant Class 1E cabling, raceways, and containment electrical penetrations (see Figure 1.9A- 6). Each redundant train has 16 circuits (eight signal and eight heater circuits).
1.9A.3.3 CORE EXIT THERMOCOUPLE SYSTEM
The core exit thermocouples provide an indication of core uncovery and clad heatup. They measure the temperature of the steam at the core exit which becomes superheated as the two-phase mixture level falls below the top of the core. The CETs provide the operator with the important information on the trend of the clad heatup.
(DRN 03-1872, R13; EC-18688 R304) Since the core exit thermocouples are an integral part of the ICI detector assemblies, their operation is dependent on that associated ICI detector string location being active. When an ICI location is made inactive by the installation of a simulated (dummy) ICI assembly, then that corresponding core exit thermocouple is inactive as well. There are currently 53 active assemblies. (DRN 03-1872, R13)
Type K (Chromel-Alumel) thermocouples are included within each of the In-Core Instrumentation (ICI) detector assemblies. The junction of each thermocouple is located above the top of the active fuel inside a tube which supports and shields the ICI detector assembly from hydraulic forces. Figure 1.9A-7 shows the axial arrangement of the CET in the calibration tube design used at Waterford 3. The core locations for the CETs are shown on Figure 1.9A-8. (EC-18688, R304) (DRN 03-2054, R14) The CETs will be qualification tested up to 1650F. Extrapolation of the data will provide calibration up to 1800F. Although the absolute accuracy is reduced above this value, the CETs continue to provide accurate information on the tread of the clad temperature. Tests have shown that the CETs continue to function up to 2300F. For the top-mounted instrumentation at Waterford 3, the thermocouples and thermocouple leads are exposed only to the core exit steam, not the higher cladding temperature. From analysis of design basis events, the maximum steam temperature is less than 1800F. (DRN 03-2054, R14) (EC-18688, R304) The external cabling, containment electrical penetrations, raceways, and signal processors of each of the ICI detector assemblies have Class 1E and seismic Category I qualification. The CETs and cabling were installed after issuance of NUREG 0578 to conform with its ICC requirement. (EC-18688, R304)
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(EC-18688, R304) Figure 1.9A-9 depicts the cable routing from the reactor vessel head to the process panels. Fifty-six CET cables (Chromel-alumel shielded twisted pairs) are installed in 56 stainless steel conduit/cable assemblies together with 56 incore neutron flux detector cables from the reactor vessel head to cable trays located in the vicinity of the reactor vessel. There is one CET' cable and one incore neutron flux detector cable in each of these conduit assemblies. The 56 CETs are then segregated at the trays located in the vicinity of the reactor vessel and have been installed in two separate trains (28 CET cables per train). Outside the containment the cables are routed via two separate trains of dedicated, independent, seismic Category I conduits up to the Class 1E signal processors in the Qualified Safety Parameter Display System (QSPDS) panels. (EC-18688, R304)
The cables and penetrations are Class 1E and have NUREG 0588 environmental qualification, also the raceways are seismic Category I. The non-safety designation of the raceways and the penetrations, as shown on Figure 1.9A-9, is due to the fact that other cables in these raceways, although identically qualified and low voltage, have non-safety instrumentation functions.
The only place where the CET cables run together with designated non-safety-related cables therefore, are in trays, conduits and penetrations (all qualified to Class 1E requirements) located inside containment. These non-safety cables are low level. Any short circuit or ground in these cables will produce a current within their current carrying capacity and therefore will not degrade the CETs signal.
(EC-18688, R304) The 28 CET cables are terminated, in each QSPDS panel, in the same manner as other Class 1E input cables. (EC-18688, R304)
1.9A.4 SIGNAL PROCESSING AND DISPLAY
In configuring the control room, it has been LP&L's design philosophy to allow as much as possible operator use of the same equipment in off-normal and emergency situations as under normal operating conditions. For this reason, primary and backup ICC display in the Waterford 3 control room has been provided for by the Qualified Safety Parameter Display System (QSPDS). The QSPDS performs safety grade signal processing and display of the ICC parameters, and is located on the main control panel for reactor protection in order to facilitate operator use. With the incorporation of access to the line printer through the plant monitoring computer, the QSPDS meets or exceeds the requirements of NUREG 0737 H.F.2 Attachment I and Regulatory Guide 1.97 for primary and backup operator displays. The QSPDS accepts sensor inputs, processes the signals, and transmits the output to its own alphanumeric display and to the plant monitoring computer. All non-Class 1E inputs and the interface with the plant monitoring computer are isolated from the Class IE QSPDS equipment. The QSPDS is capable of providing to the operator important information on the performance of many critical safety functions. However, the discussion here is centered on the processing and display of the information related to ICC and the criteria given in NUREG 0737.
(EC-18688, R304; EC-12329, R306) A spatially oriented CET temperature map is available on demand from each train of the QSPDS (primary and backup) providing a uniform representative picture of core exit temperature obtained by utilizing CETs dedicated only to that train. A recorder is provided to allow trending of representative CET temperature for the primary display (QSPDS train A). (EC-18688, R304; EC-12329, R306)
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(EC-18688, R304) Direct readout and hard copy capability is provided for all thermocouple temperatures (direct readout for the CETs associated with each train of the QSPDS can be obtained from the display associated with that train; hard copy capability via the line printer as discussed above). Selective readings of core exit temperature, continuous on demand, are available from both the primary and backup displays. (EC-18688, R304)
1.9A.4.1 QSPDS PROCESSING
The QSPDS is a two-channel, seismically qualified, Class IE system which uses a microprocessor for ICC signal processing and alphanumeric display. Each channel is electrically independent and physically separate from the other. The system is designed to achieve an availability of 99 percent.
In general, the input signal processing performed by the QSPDS consists of:
1) Checking that the sensor inputs are within their specified range.
2) Converting the sensor inputs to engineering units for display.
3) Calculating parameters from sensor inputs.
4) Calculating and initiating alarms when a parameter exceeds a setpoint.
5) Self diagnostic testing.
The QSPDS processing equipment includes operator interfaces for testing, calibration, and adjustments to be performed. In addition, automatic on-line surveillance and diagnostic test capabilities are included. These tests check for specified hardware and software malfunctions and alert the operator through the QSPDS display. It is designed to facilitate the recognition and location of the source of the malfunction to the operator.
If in the remote chance that one QSPDS channel or individual sensor fails, the operator has the following information to identify the failure:
1) QSPDS error codes and alarms.
2) Additional sensor inputs for hot leg temperature, cold leg temperature and pressurizer pressure on the control board separate from the QSPDS.
3) The HJTC and CET Systems have multiple sensors in each channel which the operator can use to correlate and check inputs.
4) The HJTC sensor output can be tested by adjusting the heater power.
The following subsections describe in more detail the processing and display for each of the ICC instruments.
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1.9A.4.1.1 Saturation Margin Monitor
The SMM/QSPDS calculates and displays both the temperature and pressure margin to saturation. The temperature saturation margin is the difference between the saturation temperature and the maximum temperature input. The pressure saturation margin is the difference between the minimum RCS pressure input and the saturation pressure. The saturation temperature is calculated using the minimum pressure input, a table of temperature and pressure values representing the saturation curve, and an interpolation routine. The saturation pressure is calculated similarly using the maximum temperature location, i.e., RTDs in the hot and cold legs, maximum of the top three unheated junction thermocouples (upper head region), and the representative CET temperature. The minimum temperature saturation margin from the RTDs and upper head is also calculated to give the operator the best indication of the RCS margin to saturation.
(DRN 03-2054, R14) An audible and display alarm is initiated when the RCS (not including CET) temperature saturation margin falls below the setpoint value of 10F subcooling. An alarm is also initiated when the CET temperature saturation margin exceeds 10F superheat. No alarms are initiated based on the pressure margin.
The following information is displayed:
Parameters Display Range
1. Temperature margin to 700F subcooled to saturation for each 2100F superheated temperature source (RTD, UHJTC, CET)
2. Pressure margin to saturation 3000 psi subcooled to for each temperature source 3000 psi superheated
3. Temperature input values RTD - 50 to 750F UHJTC - 32 to 2300F CET - 32 to 2300F
4. Pressure input value 15 to 3000 psia (DRN 03-2054, R14)
The saturation margin is identified as subcooled or superheated on the display.
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1.9A.4.1.2 Heated Junction Thermocouple System
A detailed description of the HJTC signal processing is included in CEN-185, Supplement 3 (Reference 5). A brief description of the processing performed is given below.
(DRN 03-2054, R14) 1) Determine if liquid exists at each of the HJTC sensor locations. When liquid surrounds the HJTC sensor, the differential temperature is small. When steam surrounds the sensor (i.e., the sensor is uncovered), the heated junction temperature increases and the sensor differential temperature becomes large. When the differential temperature is greater than a predetermined setpoint value (200F), the sensor is identified as being uncovered (see Figure 1.9A-10). The sensor is also identified as uncovered if the unheated junction temperature is above a setpoint value. This setpoint is high enough (700F) to ensure that steam surrounds the sensor. It is used to maintain an uncovered signal if sensor heater power is completely cut off. (DRN 03-2054, R14)
2. Calculate percent liquid level for the upper head and upper plenum regions. For the split probe design used at Waterford 3, the collapsed level in the upper head is measured independently from the collapsed level in the upper plenum. The processing and display of the collapsed level is consistent with the manner in which it is measured. That is, the percent liquid height in each region, which corresponds to the number of covered sensors in that region, is displayed separately.
3. Provide sensor heater power control signal. The sensor heater power is controlled to prevent damage to the sensor due to overheating. The input to the control logic is the maximum heated junction temperature and the maximum sensor differential temperature from all sensors. When the temperature (heated junction or differential) reaches a preset value, the power control signal is reduced linearly as a function of temperature (see Figure 1.9A-11). The minimum of the control signals derived from the heated junction and differential temperatures is used to reduce heater power. The heater power to all sensors is reduced until the temperature of the uncovered sensor stabilizers at an acceptable value. There is at all times, even when power is cutback, sufficient heater power to generate an uncovered signal if the collapsed water level falls below sensor.
4. Initiate an alarm when any HJTC sensor becomes uncovered. When any sensor differential temperature or unheated junction temperature exceeds the uncovered setpoint value, an audible and display alarm is initiated indicating that the collapsed level in the reactor vessel has decreased.
5. Perform fault condition and diagnostic testing. The system is designed to automatically detect and display several specific fault conditions associated with the HJTC instrument. These faults include open thermocouples and a loss of sensor heater power. The effect and detection of several fault conditions is discussed in Reference 7.
6. Determine the maximum unheated thermocouple temperature from the top three sensors (upper head region). The fluid temperature in the upper head is input to the SMM as described in Subsection 1.9A.4.1.2.
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The following information is displayed by the QSPDS:
Parameters Display Range
1. Percent liquid level in upper head 0 - 100%
2. Percent liquid level in upper plenum 0 - 100%
3. Status of each HJTC sensor Covered/Uncovered
(DRN 03-2054, R14) 4. Heated junction temperatures 32 – 2300F
5. Unheated junction temperatures 32 – 2300F
6. Differential temperatures -2268 - +2268F (DRN 03-2054, R14)
7. Heater power 0 - 100%
1.9A.4.1.3 Core Exit Thermocouples
(DRN 03-2054, R14; EC-18688, R304) The processing equipment for the CETs calculates the representative (maximum) CET temperature from the valid available values input to the channel. It also calculates the two highest valid CET temperatures in each quadrant. The representative CET temperature is calculated at the upper 95 percent of the distribution of valid CET temperatures with a 95 percent confidence level. CET temperatures from all four core quadrants are input and processed by each channel. These temperatures are categorized into four quadrants and identified by their location above the core. Any temperatures that are out-of-range (thereby indicating a fault) based on statistical analysis are eliminated from the calculations. The representative CET temperature is input to the SMM calculations as described in Subsection 1.9A.4.1.2. An alarm is generated when the representative CET temperature exceeds a high temperature setpoint of 670 degrees F. (DRN 03-2054, R14; EC-18688, R304)
The following information is displayed:
Parameter Display Range
(DRN 03-2054, R14) 1. Representative CET temperature 32 - 2300F
2. Two highest CET temperatures 32 - 2300F per quadrant (with identifier)
3. All CET temperatures input to 32 - 2300F channel (by quadrant with identifier) (DRN 03-2054, R14)
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1.9A.4.2 QSPDS DISPLAY
The QSPDS displays present direct, reliable, and continuous safety grade information on demand from each of the ICCI components. Existing alarm conditions and system faults are also shown. Human factors engineering is incorporated into the alphanumeric displays. Paging capabilities are provided in order to group and display all the information more efficiently.
(EC-12329, R306) The QSPDS provides Class 1E analog outputs for trending capabilities of the essential ICC detection parameters. Output from primary display (QSPDS train A) is connected to a recorder for a time history record of the ICC parameters. This trend recording aids the operator in following the progression of the ICC event. The following ICC parameters are trended: (EC-12329, R306)
1. RCS temperature saturation margin
2. CET temperature saturation margin
3. Percent liquid level in the upper head
4. Percent liquid level in the upper plenum
5. Representative CET temperature
The information from each of the ICCI components that are displayed by the QSPDS is given in Subsection 1.9A.4.1.
1.9A.5 SYSTEM VERIFICATION TESTING
This section describes the tests that have been performed to verify the ICC sensor performance. In some cases, operational experience can be used for sensor verification.
1.9A.5.1 RTD AND PRESSURIZER PRESSURE SENSORS
The hot and cold leg RTDs and the pressurizer pressure sensors are standard NSSS instruments which have a well known response. No special verification tests have been performed or are planned. These sensors provide reliable temperature and pressure inputs that are considered adequate for use in the ICCI System.
1.9A.5.2 HJTC SYSTEM
The HJTC System is a new system developed to indicate the liquid inventory (collapsed level) above the core. Since it is a new instrument system, an extensive testing program has been completed to verify its ability to indicate the liquid inventory existing above the core under conditions similar to what it may be exposed to during an ICC event.
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The test program was divided into three phases. The phase I test series consisted of feasibility and proof-of-principle tests where the concept of using HJTCs as a water level measurement device was confirmed. The Phase I results are reported in CEN-185, Supplement I (Reference 6). The Phase II tests, documented in CEN-185, Supplement 2 (Reference 3), verified the performance of a complete HJTC probe assembly under thermal-hydraulic conditions representative of what might be encountered in a PWR. Single phase, two-phase, and depressurization transients were conducted. Phase III, reported in CEN-185, Supplement 3 (Reference 5), was the final testing of a prototype HJTC System. This included verification of the integral operation of the HJTC probe assembly, sensor heater power control, and signal processing. The conclusion of these tests was that the HJTC System is capable of measuring and displaying to the operator the water inventory above the core in a reactor vessel.
In addition to the testing described above, the C-E Owners Group has completed a study to analyze the response of the HJTC System during various accidents. The purpose of the study is to quantify the effect that operation of the reactor coolant pumps during the transient has on the measured water level. A qualitative description of this effect is given in Reference 7. The results of the study also provide the basis for development of guidelines on the use of the HJTC System during accidents and in the emergency operating procedures.
1.9A.5.3 CETs
An evaluation has been conducted to verify the thermal-hydraulic performance of the CETs for use as an ICC detection sensor. This study reviewed the CET response during normal PWR operation, simulated accidents in LOFT and Semiscale tests, and during the TMI-2 accident. Analyses of CET response were also performed for conditions representative of PWR core uncovery. The evaluation concluded that the CETs are able to provide the reactor operator with information on the status and trend of fuel cladding heat-up during core uncovery.
1.9A.6 ICCI SYSTEM QUALIFICATION
The ICCI System (including the QSPDS) is environmentally and seismically qualified. The details of the qualification are provided in LP&L s "Response to Regulatory Guide 1.97, Rev. 2." The instrumentation has been seismically qualified to IEEE-344-1975.
(DRN 03-2054, R14) In addition, the HJTC System has been extensively tested and verified under conditions similar to what it may encounter during an ICC event (Reference 5). Each thermocouple is tested and calibrated up to 1200°F. Approximately one out of twenty is removed from production for testing and calibration up to 1800°F. The CETs have also been tested and verified to function up to a temperature of 2300F (Reference 8). (DRN 03-2054, R14)
1.9A.7 OPERATING INSTRUCTIONS
Guidelines for reactor operators to use to detect ICC and take corrective action have been developed by the CE Owners Group (Reference 9) and approved by the NRC for implementation (Reference 10, 11). These guidelines form the basis for the emergency operating procedures for Waterford 3 in accordance with Supplement 1 to NUREG-0737.
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1.9A.8 COMPARISON OF DOCUMENTATION REQUIREMENTS WITH THIS REPORT
Tables 1.9A-2, 1.9A-3 and 1.9A-4 provide a point by point comparison of the documentation required by Item II.F.2 of NUREG-0737, the requirements of Attachment I of Item II.F.2, and the Criteria of Appendix B of NUREG-0737 with the ICCI installed at Waterford 3.
1.9A.9 SCHEDULE FOR ICCI IMPLEMENTATION
The ICCI System was installed prior to Cycle 1. Operability of the HJTC System was deferred to Cycle 2.
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REFERENCES TO APPENDIX 1.9A
1. NUREG-0737, "Clarification of TMI Action Plan Requirements," NRC, Nov. 1980.
2. CEN-117, "Inadequate Core Cooling - A Response to NRC IE Bulletin 79-06C, Item 5 for Combustion Engineering NSSS," Combustion Engineering, Oct. 1979.
3. CEN-185, Supp. 2, "Heated Junction Thermocouple Phase II Test Report," Combustion Engineering, Nov. 1981.
4. CEN-181, "Generic Responses to NRC Questions on the CE Inadequate Core Cooling Instrumentation," Combustion Engineering, Sept. 1981.
5. CEN-185, Supp. 3, "Heated Junction Thermocouple Phase III Test Report," Combustion Engineering, Sept. 1982.
6. CEN-185, Supp. 1, "Heated Junction Thermocouple Phase I Test Report," Combustion Engineering, Nov. 1981.
7. Letter from K. Baskin (CEOG) to D. Crutchfield (NRC), June 1, 1982.
8. Anderson, R. L., Banda L. A., Cain D. G., "Incore Thermocouple Performance Under Simulated Accident Conditions," IEEE Nuclear Science Symposium, Vol. 28, 1980.
9. CEN-152, Rev. 1, "Combustion Engineering Emergency Procedure Guidelines," July, 1982.
10. Letter from D. G. Eisenhut (NRC) to R. W. Wells (CEOG), February 4, 1983.
11. Letter from J.A. Zwolinski (NRC) to R.W. Wells (CEOG), April 16, 1985.
1.9A-15
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TABLE 1.9A-1
DEFINITION OF ICC EVENT PROGRESSION INTERVALS
Interval No. ICC Phase Bounding Parameter Description
1 Approach to Reduction in RCS Depressurization of RCS Subcooling until to saturation pressure of saturation occurs hot leg temperature or heatup to saturation temperature at safety valve pressure.
2 Approach to Falling collapsed Loss of coolant inventory water level above from RCS with boiling core from continued depressurization and/or decay heat.
3 Approach to Two-phase mixture Core uncovery causes clad and/or existence level falls below heatup and production of of top of core superheated steam at core resulting in clad exit. heatup
4 Recovery from Two-phase level Coolant addition by ECCS rises above top raises water level and of core as RCS cools fuel. Recovery refills from ICC is complete when reactor vessel is full or when stable, controllable conditions exist.
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TABLE 1.9A-2
COMPARISON OF ICCI TO DOCUMENTATION REQUIREMENTS
OF ITEM II.F.2 OF NUREG-0737
Item Response la. A description of the ICCI System is provided in Section 1.9A.3. New instrumentation that was added included the HJTC probe assemblies. Display of the ICC parameters will be on the QSPDS. lb. Existing instrumentation which can aid the operators in the detection of ICC is discussed in Reference 2. Waterford 3 will use the SMM and CETs as part of the ICCI System. lc. The final ICCI System is as described in Section 1.9A.3 and 1.9A.4.
2. The design analysis and testing performed to evaluate the ICCI is discussed in Reference 2 and Section 1.9A.5.
3. Additional instrumentation testing is discussed in Section 1.9A.5. System qualification testing is discussed in Section 1.9A.6.
4. This table evaluates the ICCI conformance to Item II.F.2 of NUREG-0737. Table 1.9A-3 evaluates conformance to Attachment 1 of Item II.F.2. Table 1.9A-4 evaluates conformance to Appendix B of NUREG-0737.
5. Section 1.9A.4 describes the processing and display of the ICC parameters which is incorporated into the QSPDS.
6. Section 1.9A.9 discusses the schedule for installation and implementation of the ICCI System.
7. Guidelines for use of the ICCI are discussed in Section 1.9A.7. An ICCI functional description is given in Section 1.9A.2. 8. Section 1.9A.7 discusses the current emergency operating procedures.
9. LP&L has made all submittals necessary for Item II.F.2.
WSES-FSAR-UNIT-3
TABLE 1.9A-3 Revision 306 (05/12)
COMPARISON OF ICCI TO ATTACHMENT 1 OF II.F.2
Item Response
1. Waterford 3 has 56 CETs distributed uniformly across the top of the core. Figure 1.9A-8 shows the locations of the CETS.
2a. A spatial CET temperature map is available on demand from the QSPDS display.
2b. The maximum CET temperature calculated from a statistical analysis, is used as a representative temperature and is displayed continuously on demand.
2c. The QSPDS provides direct readout of the CET temperatures. The line printer provides hard-copy recording. The display range is from 32°F to 2300°F.
(EC-12329, R306) 2d. Trending of the representative CET temperature is provided by an analog output from the QSPDS to a recorder. (EC-12329, R306)
2e. The QSPDS provides visual alarm capability as well as output to the plant annunciator for audible alarms.
2f. The QSPDS incorporates human factors engineering.
3. The QSPDS meets, being a redundant system, the requirements for a safety grade backup display system. Both channels together display all CET temperatures.
4. The QSPDS design incorporates human factors engineering in determining the types and locations of displays and alarms. The use of these displays will be addressed in operating procedures, emergency procedures, and operator training.
5. The ICCI is evaluated for conformance to Appendix B in Table 1.9A-4.
6. The QSPDS is an electrically independent Class 1E System. It meets the applicable display requirements as modified by NUREG-0737, Supplement 1.
7. The ICCI is qualified as described in Section 1.9A.6.
8. The QSPDS is designed to provide an availability of 99 percent. Availability of the ICCI is addressed in the Technical Specifications.
9. The quality assurance provision of Appendix B, Item 5, will be applied to the ICCI.
WSES-FSAR-UNIT-3
TABLE 1.9A-4 (Sheet 1 of 2) Revision 306 (05/12)
COMPARISON TO ICCI TO APPENDIX B OF NUREG-0737
Item Response
1. The ICCI is environmentally and seismically qualified as described in Section 1.9A.6. The isolation devices in the QSPDS are accessible for maintenance.
2. The ICCI through the QSPDS display meets the single failure requirement. If one channel should fail the self diagnostic capability of the QSPDS, as well as additional sensor displays aid the operator in determining which channel may have failed (Subsection 1.9A.4.1).
3. The ICCI through the QSPDS is powered by Class 1E power sources.
4. Availability of the ICCI is addressed in the Technical Specifications.
5. The ICCI through the QSPDS meets quality assurance requirements for Class 1E equipment. This item was addressed in the response to Regulatory Guide 1.97, Rev 2.
6. The QSPDS provides continuous displays on demand.
(EC-12329, R306)
7. The QSPDS provides trend recording with a recorder. (EC-12329, R306)
8. The QSPDS displays are clearly identified on the control panel and are human factor engineered.
9. Output signals from the QSPDS to non-qualified equipment are transmitted through Class 1E isolation devices.
10. The operational availability of the ICCI can be checked as described in Subsection 1.9A.4.1.
11. Servicing, testing, and calibration programs for the ICCI through the QSPDS shall be specified in plant operating procedures.
12. The means for removing channels from service have been considered in the ICCI design.
WSES-FSAR-UNIT-3
TABLE 1.9A-4 (Sheet 2 of 2)
COMPARISON TO ICCI TO APPENDIX B OF NUREG-0737
Item Response
13. The design facilitates administrative control of access to all setpoint adjustments, calibration adjustments, and test points.
14. The design minimizes anomalous indications which might confuse operator.
15. The design facilitates the recognition, location, replacement, repair or adjustment of malfunctioning components.
16. The design directly measures the desired variables to the extent practical.
17. The design incorporates this requirement to the extent practical. 18. Periodic testing of the instrument channels has been incorporated.
WSES-FSAR-UNIT-3
APPENDIX 1A
FSAR ACRONYMS
ACB AIR CIRCUIT BREAKER
ACCWS AUXILIARY COMPONENT COOLING WATER SYSTEM
ACCW AUXILIARY COMPONENT COOLING WATER
ACPI AUXILIARY CONTROL PANEL INSTRUMENTATION
ADS AUTOMATIC DISPATCHING SYSTEM
ADT AVERAGE DAILY TRAFFIC
AGA AUTOMATIC GAS ANALYZER
AH AIR HANDLING UNIT
AHP ABOVE HEAD OF PASSES
ALARA AS LOW AS REASONABLY ACHIEVABLE
AMI AUTOMATIC - CEA - MOTION INHIBIT
APD AMPLITUDE PROBABILITY DISTRIBUTION
AP/SR ACTIVE TRIAXIAL PEAK SHOCK RECORDER
ARRS AIRBORNE RADIOACTIVITY REMOVAL SYSTEM
ASI AXIAL SHAPE INDEX
ATWS ANTICIPATED TRANSIENTS WITHOUT SCRAM
AWP AUTOMATIC - CEA - WITHDRAWAL PROHIBIT
BAMT BORIC ACID MAKEUP TANKS
BMS BORON MANAGEMENT SYSTEM
BOC BEGINNING OF CYCLE
BOL BEGINNING OF LIFE
BOP BALANCE OF PLANT
BWR BOILING WATER REACTOR
CAMS CHEMICAL ADDITION METERING SYSTEM
1A-1
WSES-FSAR-UNIT-3
APPENDIX 1A
FSAR ACRONYMS (Cont’d)
CARS CONTAINMENT ATMOSPHERE RELEASE SYSTEM
ASME AMERICAN SOCIETY OF MECHANICAL ENGINEERS
ANSI AMERICAN NATIONAL STANDARDS INSTITUTE
ASTM AMERICAN SOCIETY OF TESTING MATERIALS
CCS CONTAINMENT COOLING SYSTEM
CCWS COMPONENT COOLING WATER SYSTEM
CEA CONTROL ELEMENT ASSEMBLY
CEAC CEA CALCULATOR
CEADS CONTROL ELEMENT ASSEMBLY DRIVE SYSTEM
CEDM CONTROL ELEMENT DRIVE MECHANISM
CEDMCS CONTROL ELEMENT DRIVE MECHANISM CONTROL SYSTEM
CET CORE EXIT THERMOCOUPLE
CF CONCENTRATION FACTOR
CFWS CONDENSATE AND FEEDWATER SYSTEMS
CGCS COMBUSTIBLE GAS CONTROL SYSTEM
CHF CRITICAL HEAT FLUX
CIAS CONTAINMENT ISOLATION ACTUATION SIGNAL
CIS CONTAINMENT ISOLATION SYSTEM
CIV CONTAIMENT ISOLATION VALVE(S)
COLSS CORE OPERATING LIMIT SUPERVISORY SYSTEM
CPC CORE PROTECTION CALCULATORS
CPIS CONTAINMENT PURGE ISOLATION SIGNAL
CSAS CONTAINMENT SPRAY ACTUATION SIGNAL
CSB CORE SUPPORT BARREL
CSP CONDENSATE STORAGE POOL
1A-2
WSES-FSAR-UNIT-3
APPENDIX 1A
FSAR ACRONYMS (Cont’d)
CSS CONTAINMENT SPRAY SYSTEM
C/U CONTROL UNIT
CVAS CONTROLLED VENTILATION AREA SYSTEM
CVCS CHEMICAL AND VOLUME CONTROL SYSTEM
CVH CONTAINMENT VENT HEADER
CWS CIRCULATING WATER SYSTEM
DBA DESIGN BASIS ACCIDENT
DBE DESIGN BASIS EARTHQUAKE
DE DOUBLE ENDED
DEH DIGITAL ELECTRO-HYDRAULIC
DF DECONTAMINATION FACTOR
(DRN 06-623, R15) DLR DOSIMETER OF LEGAL RECORD (DRN 06-623, R15)
DNB DEPARTURE FROM NUCLEATE BOILING
DNBR DEPARTURE FROM NUCLEATE BOILING RATIO
DWS DEMINERALIZED WATER SYSTEM
EAB EXCLUSION AREA BOUNDARY
ECCS EMERGENCY CORE COOLING SYSTEM
EFAS EMERGENCY FEEDWATER ACTUATION SIGNAL
EFPD EFFECTIVE FULL POWER DAYS
EFPY EFFECTIVE FULL POWER YEARS
EFS EMERGENCY FEEDWATER SYSTEM
1A-3 Revision 15 (03/07)
WSES-FSAR-UNIT-3
APPENDIX 1A
FSAR ACRONYMS (Cont’d)
EFW EMERGENCY FEEDWATER SYSTEM
EOC END OF CYCLE
ESF ENGINEERED SAFETY FEATURES
ESFAS ENGINEERED SAFETY FEATURES ACTUATION SYSTEM
FHB FUEL HANDLING BUILDING
FHBVS FUEL HANDLING BUILDING VENTILATION SYSTEM
FMEA FAILURE MODES AND EFFECTS ANALYSIS
FSAR FINAL SAFETY ANALYSIS REPORT
FPS FUEL POOL SYSTEM
FWCS FEEDWATER CONTROL SYSTEM
GSH GAS SURGE HEADER
GTA GAS TUNGSTEN ARC
GWMS GASEOUS WASTE MANAGEMENT SYSTEM
HEPA HIGH EFFICIENCY PARTICULATE AIR
HP HIGH PRESSURE
HPSI HIGH PRESSURE SAFETY INJECTION
HRS HYDROGEN RECOMBINER SYSTEM
HVAC HEATING VENTILATION, AND AIR CONDITIONING
ICC INADEQUATE CORE COOLING
ILRT INTEGRATED LEAK RATE TEST
IPARMS IN PLANT AREA RADIATION MONITORING SYSTEM
ISEG INDEPENDENT SAFETY ENGINEERING GROUP (EC-14275, R306) ISFSI INDEPENDENT SPENT FUEL STORAGE INSTALLATION (EC-14275, R306) (DRN 00-576, R11) ITR INDEPENDENT TECHNICAL REVIEW (DRN 00-576, R11)
JTG JOINT TEST GROUP
LCO LIMITING CONDITION FOR OPERATION
1A-4 Revision 306 (05/12)
WSES-FSAR-UNIT-3
APPENDIX 1A
FSAR ACRONYMS (Cont’d)
LDA LOAD DROP ANTICIPATION
LMFW LOSS OF MAIN FEEDWATER
LOCA LOSS OF COOLANT ACCIDENT
LOCV LOSS OF CONDENSER VACUUM
LOOP LOSS OF OFFSITE POWER
LOOP LOUISIANA OFFSHORE OIL PORT
LP LOW PRESSURE
LPD LOCAL POWER DENSITY
LPSI LOW PRESSURE SAFETY INJECTION
LPZ LOW POPULATION ZONE
LSE LEAD STARTUP ENGINEER
LSSS LIMITING SAFETY SYSTEM SETTING
LTC LONG TERM COOLING
LTOP LOW TEMPERATURE OVERPRESSURE PROTECTION
LWMS LIQUID WASTE MANAGEMENT SYSTEM
MCC MOTOR CONTROL CENTER
MCES MAIN CONDENSER EVACUATION SYSTEM
MCR MAIN CONTROL ROOM
MDP MOTOR DRIVEN PUMP
MFIV MAIN FEEDWATER ISOLATION VALVE
1A-5 Revision 11 (5/01)
WSES-FSAR-UNIT-3
APPENDIX 1A
FSAR ACRONYMS (Cont’d)
MFWLB MAIN FEEDWATER LINE BREAK
MFWS MAIN FEEDWATER SYSTEM
MM MODIFIED MERCALLI
MPC MAXIMUM PERMISSIBLE CONCENTRATION
MR&T MISSISSIPPI RIVER & TRIBUTARIES
MS MAIN STEAM SUPPLY SYSTEM
MSIS MAIN STEAM ISOLATION SIGNAL
MSIV MAIN STEAM ISOLATION VALVE
MSL MEAN SEA LEVEL
MSLB MAIN STEAM LINE BREAK
MSR MAXIMUM STEAMING RATE
MTC MODERATOR TEMPERATURE COEFFICIENT
M&TE MEASURING & TEST EQUIPMENT
M/TR MAGNETIC TAPE RECORDER
MYBP MILLION YEARS BEFORE PRESENT
NBS NATIONAL BUREAU OF STANDARDS
NNS NON-NUCLEAR SAFETY
NPIS NUCLEAR PLANT ISLAND STRUCTURE
NPSH NET POSITIVE SUCTION HEAD
NRC NUCLEAR REGULATORY COMMISSION
NRR NUCLEAR REACTOR REGULATION
NSSS NUCLEAR STEAM SUPPLY SYSTEM
1A-6 Revision 11 (5/01)
WSES-FSAR-UNIT-3
APPENDIX 1A
FSAR ACRONYMS (Cont’d)
OBE OPERATING BASIS EARTHQUAKE
OEEG OPERATIONAL ENGINEERING EXPERIENCE GROUP
OPC OVERSPEED PROTECTION CONTROLLER
OL OPERATING LICENSE (DRN 03-657, R12-C) OSRC ON-SITE SAFETY REVIEW COMMITTEE (DRN 03-657, R12-C) P/A TRIAXIAL PEAK ACCELEROGRAPH
PABX PRIVATE AUTOMATIC BRANCH EXCHANGE
PAMI POST-ACCIDENT MONITORING INSTRUMENTATION
PCS PLANT COMPUTER SYSTEM
PDF PROJECT DESIGN FLOOD
PDIL POWER DEPENDENT INSERTION LIMIT
PLCEA PART LENGTH CONTROL ELEMENT ASSEMBLY
PMF PROBABLE MAXIMUM FLOOD
PMH PROBABLE MAXIMUM HURRICANE
PMP PROBABLE MAXIMUM PRECIPITATION
PORC PLANT OPERATIONS REVIEW COMMITTEE
PORV POWER OPERATED RELIEF VALVE
PPS PLANT PROTECTION SYSTEM
PP/SR PASSIVE TRIAXIAL PEAK SHOCK RECORDER
P/SA TRIAXIAL PEAK SHOCK ANNUNCIATOR
PSAR PRELIMINARY SAFETY ANALYSIS REPORT
PSD POWER SPECTRAL DENSITY
P/SR PLAYBACK UNIT STRIP RECORDER
1A-7 Revision 12-C (07/03)
WSES-FSAR-UNIT-3
APPENDIX 1A
FSAR ACRONYMS (Cont’d)
PSS PRIMARY SAMPLING SYSTEM
PSWS POTABLE AND SANITARY WATER SYSTEM
PVMF PRECRITICAL VIBRATION MOAITORING PROGRAM
PWR PRESSURIZED WATER REACTOR
PWHT POST WELD HEAT TREATMENT
PWSS PRIMARY WATER STORAGE SYSTEM
PWTP PRIMARY WATER TREATMENT PLANT
QA QUALITY ASSURANCE
QAC QUALITY ASSURANCE COMMITTEE (DRN 99-1093) QAI QUALITY ASSURANCE INSPECTION (DRN 99-1093) QC QUALITY CONTROL
QI QUALITY INSTRUCTION
QP QUALITY PROCEDURE
QR QUALITY REQUIREMENT
QSPDS QUALIFIED SAFETY PARAMETER DISPLAY SYSTEM
RAB REACTOR AUXILIARY BUILDING
RAP RESTRAIAT ACTUATION PROGRAM
RAS RECIRCULATION ACTUATION SIGNAL
RCGVS REACTOR COOLANT GAS VENTING SYSTEM
RCP REACTOR COOLANT PUMP
RCPB REACTOR COOLANT PRESSURE BOUNDARY
RCS REACTOR COOLANT SYSTEM
RDT REACTOR DRAIN TANK
1A-8 Revision 11 (5/01)
WSES-FSAR-UNIT-3
APPENDIX 1A
FSAR ACRONYMS (Cont’d)
RMS RADIATION MONITORING SYSTEM
RPCCP REACTOR POWER CUTBACK COATROL PANEL
RPCM REACTOR POWER CUTBACK MODULE
RPCS REACTOR POWER CUTBACK SYSTEM
RPS REACTOR PROTECTIVE SYSTEM
RRS REACTOR REGULATING SYSTEM
RSPT REED SWITCH POSITION TRANSMITTER
RTD RESISTANCE TEMPERATURE DETECTOR
RTGB REACTOR TURBINE GENERATOR BOARD
RTSG REACTOR TRIP SWITCHGEAR
RV REACTOR VESSEL
RLIS REFUELING LEVEL INDICATING SYSTEM
RWLIS REFUELING WATER LEVEL INDICATING SYSTEM
RWSP REFUELING WATER STORAGE POOL
SA SPECTRAL ANALYSIS
SAFDL SPECIFIED ACCEPTABLE FUEL DESIGN LIMITS
SB STATION BLACKOUT
SBS STEAM BYPASS SYSTEM
SBCS STEAM BYPASS CONTROL SYSTEM
SBVS SHIELD BUILDING VENTILATION SYSTEM
SDCHX SHUTDOWN COOLING HEAT EXCHANGER
SDCS SHUTDOWN COOLING SYSTEM
SER SAFETY EVALUATION REPORT
1A-9 Revision 11 (5/01)
WSES-FSAR-UNIT-3
APPENDIX 1A
FSAR ACRONYMS (Cont’d)
SFP SPENT FUEL POOL
SG STEAM GENERATOR
SGBDS STEAM GENERATOR BLOWDOWN DEMINERALIZER SYSTEM
SGBS STEAM GENERATOR BLOWDOWN SYSTEM
SGTR STEAM GENERATOR TUBE RUPTURE
SIAS SAFETY INJECTION ACTUATION SIGNAL
SIS SAFETY INJECTION SYSTEM
SISC SAFETY INJECTION SYSTEM COMPONENTS
SIT SAFETY INJECTION TANK
SMSA STANDARD METROPOLITAN STATISTICAL AREA
SPCS STEAM AND POWER CONVERSION SYSTEM
SPDS SAFETY PARAMETER DISPLAY SYSTEM
SPS STANDARD PROJECT STORM
SR SURVEILLANCE REQUIREMENT
SRC SAFETY REVIEW COMMITTEE
SRO SENIOR REACTOR OPERATOR
SRSS SQUARE ROOT OF THE SUM OF THE SQUARES
S/S SEISMIC SWITCH
SS/CU SEISMIC SWITCH CONTROL UNIT
SSE SAFE SHUTDOWN EARTHQUAKE
SSS SECONDARY SAMPLING SYSTEM
STA SHIFT TECHNICAL ADVISOR
S/U STARTER UNIT
1A-10 Revision 11 (5/01)
WSES-FSAR-UNIT-3
APPENDIX 1A
FSAR ACRONYMS (Cont’d)
SWANS SOLID WASTE MANAGEMENT SYSTEM
T/A TRIAXIAL TIME-HISTORY ACCELEROGRAPH
TB TURBINE BUILDING
TCCWS TURBINE CLOSED COOLING WATER SYSTEM
TCS TURBINE CONTROL SYSTEM
TD THEORETICAL DENSITY
TDP TURBINE DRIVEN PUMP
TGSS TURBINE GLAND SEALING SYSTEM
TLD THERMOLUMINESCENT DOSIMETER
TSC TECHNICAL SUPPORT CENTER
TSP TRISODIUM PHOSPHATE DODECAHYDRATE
TSSV TURBINE STEAM SUPPLY VALVES
UGS UPPER GUIDE STRUCTURE
VCT VOLUME CONTROL TANK
VGCH VENT GAS COLLECTION HEADER
V&LPMS VIBRATION & LOOSE PARTS MONITORING SYSTEM
WMS WASTE MANAGEMENT SYSTEM
WSG WATERFORD 3 STARTUP GROUP
YBP YEARS BEFORE PRESENT
1A-11 Revision 11 (5/01)